JP2004028903A - Device for separating radioactive waste liquid and disposal system for radioactive ion exchange resin provided with the same - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To stabilize a solidified body formed by caking radioactive waste liquid. <P>SOLUTION: This device for separating radioactive waste liquid is provided with a concentrated waste liquid tank 77 which stores radioactive waste liquid containing sodium sulfate, an extracting pump 10 which extracts the supernatant of the stored radioactive waste liquid, an inorganic ion adsorbent tower 12 which removes a radioactive substance from the supernatant by making the radioactive substance contained in the the supernatant extracted adsorb on an absorbent and generates low radioactivity waste liquid whose the radiation concentration is lower than that of the radioactive waste liquid, a cement solidifying device 16 which solidifies the low radioactivity waste liquid generated by the inorganic ion adsorbent tower 12, a buffer tank 14 which separates and recovers the deposition by adding water to the deposition of the radioactive waste liquid stored in the concentrated waste liquid tank 77, and a cement solidifying device 18 which solidifies a high radioactivity mixture body whose the radiation concentration is higher than that of radioactive waste liquid formed by mixing the deposition recovered by the buffer tank 14 with the adsorbent in which the radioactive substance is adsorbed in the inorganic ion adsorbent tower 12. <P>COPYRIGHT: (C)2004,JPO

Description

[0001]
TECHNICAL FIELD OF THE INVENTION
The present invention provides, for example, a treatment system for discarding a radioactive ion exchange resin that has adsorbed radioactive ions contained in system water containing radionuclides of a nuclear power plant, and a radioactive waste liquid containing sodium sulfate obtained by the treatment system. The present invention relates to a radioactive waste liquid separation apparatus for separating into a radioactive waste liquid having a low activity concentration and a radioactive waste solution having a low sodium sulfate concentration and a high activity concentration.
[0002]
[Prior art]
In a nuclear facility such as a nuclear power plant, an ion exchange resin is used to adsorb radioactive ions. Specifically, it is used to adsorb radioactive ions contained in radionuclide-containing system water of a nuclear power plant and to purify radionuclide-containing system water. With the operation of a nuclear power plant, spent radioactive ion exchange resin that has absorbed radioactive ions is generated as waste.
[0003]
The ion exchange resin is generally based on a copolymer of styrene and divinylbenzene (D.V.B), and a sulfonic acid group in the case of a cation exchange resin as an ion exchange group, and an anion exchange resin. Is an aromatic organic polymer compound having a structure in which a quaternary ammonium group is bonded.
[0004]
The system water containing radionuclides includes radioactive intensities such as Co-60 (cobalt 60), Cs-137 (cesium 137), Sr-90 (strontium 90) and other Fe (iron) and Ni (nickel). And anionic species having low radioactivity such as organic substances containing I (iodine) and C-14 (carbon 14). Therefore, cation species such as Co-60, Cs-137 and Sr-90 in the system water containing radionuclides are adsorbed on the cation exchange resin, and anion species such as organic substances containing I and C-14 are It is purified by being adsorbed on an anion exchange resin.
[0005]
The ion exchange resin has a radioactive ion exchange resin tower filled with almost the same amount of cation exchange resin and anion exchange resin, and has the ability to remove cation or anion species from system water (decontamination coefficient). Is reduced to a predetermined capacity or less, the used radioactive ion exchange resin is discarded, and the radioactive ion exchange resin tower is filled with unused cation exchange resin and anion exchange resin. The volume ratio of radioactive ion exchange resin treated as waste accounts for only a small percentage of the total waste generated at nuclear power plants, but the amount of radioactivity due to substances adsorbed by the radioactive ion exchange resin is small. The total accounts for the majority of the radioactivity in waste generated during normal operation of nuclear power plants. As described above, most of the radioactivity contained in the radioactive waste discharged from the nuclear power plant during operation is adsorbed by the radioactive ion exchange resin.
[0006]
When such a used radioactive ion exchange resin is treated as waste, it is performed as follows. That is, when it is confirmed that the decontamination coefficient of the cation exchange resin and the anion exchange resin filled in the radioactive ion exchange resin tower has decreased to a predetermined value or less, as shown in FIG. It is transferred from a resin tower (not shown) to a used resin storage tank 62, where it is stored for a while. During this storage period, nuclides having a short half-life such as Co-58 as a cation and I-131 as an anion can be attenuated.
[0007]
As described above, when the nuclide with a short half-life has attenuated, the used radioactive ion exchange resin is temporarily transferred from the used resin storage tank 62 to the waste resin supply tank 63 and further transferred to the eluator 64 therefrom. It is. An aqueous sulfuric acid solution is passed through the used radioactive ion exchange resin transferred to the eluator 64. Thereby, the radioactive ions adsorbed on the used radioactive ion exchange resin migrate into the aqueous sulfuric acid solution and are eluted.
[0008]
The ion exchange resin from which the radioactive substance has been eluted and removed in this way is taken out from the lower part of the eluator 64, transferred to the incinerator 68 by the eluted resin transport container 67, and incinerated there.
[0009]
On the other hand, radioactive ions are dissolved in the aqueous sulfuric acid solution passed through the radioactive ion exchange resin in the eluator 64. This aqueous sulfuric acid solution is introduced from the eluator 64 into the left chamber 70 of the sulfuric acid recovery unit 69. In the sulfuric acid aqueous solution, a radioactive substance and sulfuric acid are separated by a diffusion dialysis membrane 71 provided in the sulfuric acid recovery device 69. That is, as shown in FIG. 4, of the sulfuric acid aqueous solution introduced from the eluator 64 into the left chamber 70, the sulfuric acid is diffused and recovered to the right chamber 72 through the diffusion dialysis membrane 71, while the radioactive substance is left. Remains in room 70. The sulfuric acid collected on the right chamber 72 side is sent to the eluator 64 and reused. As a result, a waste liquid containing a radioactive substance is obtained in the left chamber 70.
[0010]
As shown in FIG. 5, the waste liquid in the early stage of elution obtained in the left chamber 70 (hereinafter, referred to as “A-mode waste liquid”) is the late waste liquid obtained in the left chamber 70 (hereinafter, “B-mode waste liquid”). Radioactivity concentration is significantly higher.
[0011]
Therefore, the A-mode waste liquid is neutralized by adding NaOH as a neutralizing agent supplied from the neutralizing agent tank 75 in the first neutralizing tank 73 (# 1), and sulfuric acid as a neutralized body is added. After sodium (Na ₂ SO ₄ ) is produced, it is sent to the evaporator 76. Then, after being evaporated and concentrated by the evaporator 76, it is sent to the concentrated waste liquid tank 77 where it is stored for a long time. This causes the cement to solidify after the radioactivity has decayed. Such evaporative concentration is important from the viewpoint of reducing the volume of the A-mode waste liquid and reducing the amount of waste generated.
[0012]
On the other hand, the B-mode waste liquid is transferred to the neutralization tank 73 (# 2) due to the low radioactivity concentration, and here, similarly, NaOH, which is the neutralizing agent supplied from the neutralizing agent tank 75, is added. To produce a neutralized form of sodium sulfate (Na ₂ SO ₄ ). After that, it is sent to the liquid waste treatment system 78 and solidified by asphalt or the like.
[0013]
[Problems to be solved by the invention]
However, such a conventional method for treating a radioactive ion exchange resin has the following problems.
[0014]
Since the A-mode waste liquid contains sodium sulfate, when the A-mode waste liquid is concentrated, sodium sulfate is also concentrated at the same time, and the concentration of the contained sodium sulfate increases. It is known that when cement contains sodium sulfate therein, it forms ettringite, and damages such as cracks occur as the concentration of sodium sulfate contained increases. In order to bury a solidified cement with high radioactivity and high sodium sulfate concentration, it is necessary to strengthen the barrier function for radionuclides in the burial facility.
[0015]
The present invention has been made in view of such circumstances.Sodium sulfate is put into a solidified substance having a low radioactivity concentration, buried and disposed in a facility for a low radioactivity concentration, and is transferred to a facility for a high radioactivity concentration. The overall cost is reduced by burying as little sodium sulfate as possible. That is, the A-mode waste liquid is separated into a low-activity waste liquid having a lower radioactivity concentration than the A-mode waste liquid and a high-activity mixture having a lower sodium sulfate concentration and a higher radioactivity concentration than the A-mode waste liquid. By solidifying and disposing each solidified body in a separate disposal facility according to the level of radioactivity contained, the barrier function for radionuclides in the disposal facility for high radioactivity concentration is not strengthened Another object of the present invention is to provide a radioactive waste liquid separation device capable of reducing the cost as a whole, and a radioactive ion exchange resin treatment system including the same.
[0016]
[Means for Solving the Problems]
In order to achieve the above object, the present invention takes the following measures.
[0017]
That is, the radioactive waste liquid separation device according to the first aspect of the present invention extracts a waste liquid tank for storing a radioactive waste liquid containing sodium sulfate (Na ₂ SO ₄ ) and a supernatant of the radioactive waste liquid stored in the waste liquid tank. An extraction pump and an adsorbent are provided inside, and radioactive substances contained in the supernatant liquid extracted by the extraction pump are removed from the supernatant liquid by adsorbing the radioactive substances to the adsorbent, so that the radioactive waste liquid is removed more than the radioactive waste liquid. A radioactive substance removing means for producing a low radioactive waste liquid having a low radioactivity concentration; a first solidifying means for solidifying the low radioactive waste liquid generated by the radioactive substance removing means; and a radioactive waste liquid stored in a waste liquid tank. A sediment collecting means for separating and collecting the sediment by adding water to the sediment; a sediment collected by the sediment collecting means; Second solidifying means for solidifying a high-activity mixture having a higher activity concentration than the radioactive waste liquid obtained by mixing the adsorbent with the adsorbed substance.
[0018]
The system for treating a radioactive ion exchange resin according to the second aspect of the present invention is to pass at least an aqueous sulfuric acid solution for elution through the ion exchange resin to which the radioactive ion species are adsorbed, and to elute the radioactive ion species adsorbed to the ion exchange resin A sodium hydroxide (NaOH) is added as a neutralizing agent to an eluted aqueous sulfuric acid solution obtained by eluting a radioactive ion species into the aqueous sulfuric acid solution for elution, and a neutralized sulfuric acid. In a radioactive ion exchange resin treatment system, comprising: a neutralizing means for generating a neutralized waste liquid containing sodium (Na ₂ SO ₄ ); and a concentrating means for concentrating the neutralized waste liquid to generate a concentrated neutralized waste liquid. The following is added.
[0019]
That is, a waste liquid tank for storing the concentrated neutralized waste liquid generated by the concentration means, an extraction pump for extracting the supernatant of the concentrated neutralized waste liquid stored in the waste liquid tank, and an adsorbent therein, The radioactive substance contained in the supernatant extracted by the extraction pump is absorbed by an adsorbent to remove the radioactive substance from the supernatant, thereby producing a low-radioactive waste liquid with a lower radioactivity concentration than the concentrated neutralized waste liquid. Radioactive substance removing means, a first solidifying means for solidifying the low radioactive waste liquid generated by the radioactive substance removing means, and water by adding water to the precipitate of the concentrated neutralized waste liquid stored in the waste liquid tank. A sediment collecting means for separating and collecting the sediment, a sediment collected by the sediment collecting means, and an adsorbent to which the radioactive substance is adsorbed in the radioactive substance removing means during concentration. The radioactive waste separator device and a second solidification means for solidifying the high radioactivity concentrations highly radioactive mixture than waste.
[0020]
Therefore, in the first and second aspects of the present invention, by taking the above measures, the A-mode waste liquid which is a radioactive waste liquid containing sodium sulfate (Na ₂ SO ₄ ) is radiated more than the A-mode waste liquid. It can be separated into a low radioactive waste liquid having a low active concentration and a high radioactive mixture having a low sodium sulfate concentration and a high radioactive concentration than the A-mode waste liquid, and each can be solidified.
[0021]
Furthermore, the solidified body obtained by solidifying the solidified body obtained by solidifying the low-radioactive waste liquid and the solidified body obtained by solidifying the high-radioactive mixture are disposed of in separate disposal facilities. By doing so, it is possible to reduce the overall cost without strengthening the barrier function for radionuclides in the buried disposal facility for high radioactivity concentration.
[0022]
BEST MODE FOR CARRYING OUT THE INVENTION
Hereinafter, embodiments of the present invention will be described with reference to the drawings.
[0023]
In the drawings used in the following description of the embodiment, the same parts as those in FIGS. 3 and 4 are denoted by the same reference numerals and the description thereof will be omitted, and only different parts will be described here.
[0024]
An embodiment of the present invention will be described with reference to FIGS.
[0025]
FIG. 1 is a system configuration diagram showing an example of a radioactive ion exchange resin treatment system to which a radioactive waste liquid separation device according to an embodiment of the present invention is added.
[0026]
That is, the radioactive ion exchange resin treatment system to which the radioactive waste liquid separation device according to the embodiment of the present invention is added is different from the conventional radioactive ion exchange resin treatment system shown in FIG. The apparatus further includes a radioactive waste liquid separating device including a chemical tower 12, a pure water supply system 13, a batch tank 14, a cement solidifying device 16, and a cement solidifying device 18.
[0027]
The concentrated waste liquid tank 77 stores an A-mode concentrated waste liquid containing radioactive substances such as Co (cobalt), Cs (cesium), Sr (strontium), and Nb (niobium), and sodium sulfate. The A-mode concentrated waste liquid is separated into a supernatant liquid and a coagulated sediment in the concentrated waste liquid tank 77, and most of the sodium sulfate is dissolved and contained in the supernatant liquid. On the other hand, most of the radioactive substances that become hydroxides form aggregate precipitates.
[0028]
The supernatant extraction pump 10 extracts the supernatant of the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77, and sends the extracted supernatant to the inorganic ion adsorbent tower 12.
[0029]
The inorganic ion adsorbent tower 12 includes an adsorbent therein, and is contained in the supernatant liquid by the supernatant liquid being sent from the lower part in FIG. 1 to the upper part in FIG. 1 by the supernatant liquid extraction pump 10. The radioactive substance is adsorbed on the adsorbent. The supernatant liquid is sent to the cement solidifying device 16 after the radioactive substance is removed in the inorganic ion adsorbent tower 12 and the radioactivity concentration is reduced in this way. On the other hand, when the adsorbent adsorbs the radioactive substance contained in the supernatant, the adsorbent is discharged to the cement solidifying device 18 as waste.
[0030]
The pure water supply system 13 supplies pure water in an amount corresponding to the extracted supernatant liquid into the concentrated waste liquid tank 77 in which the supernatant liquid is extracted and the coagulated sediment is left. Thereby, the sodium sulfate remaining in the concentrated waste liquid tank 77 as the coagulated sediment is diluted together with the coagulated sediment.
[0031]
The batch tank 14 is supplied with pure water by the pure water supply system 13 in the concentrated waste liquid tank 77, and receives diluted coagulated sediment from the concentrated waste liquid tank 77. Then, the received coagulated sediment is stored until it is cement-solidified by the cement-solidifying device 18.
[0032]
The cement solidifying device 16 solidifies the supernatant liquid sent from the inorganic ion adsorbent tower 12 with cement. This supernatant has a sodium sulfate concentration similar to that of the A-mode concentrated waste liquid stored in the concentrated waste tank 77, but has a lower radioactivity concentration. Therefore, the solidified cement solidified by the cement solidifying device 16 has a lower radioactivity concentration than the solidified cement obtained by solidifying the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77 as it is, so that it can be easily buried. Since it can be disposed of at a disposal facility, it can be disposed at low cost.
[0033]
The cement solidifying device 18 receives the diluted coagulated sediment from the batch tank 14 and the adsorbent adsorbing the radioactive substance from the inorganic ion adsorbent tower 12 and solidifies them. The diluted coagulated sediment has a much lower sodium sulfate concentration than the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77. On the other hand, the adsorbent that has absorbed the radioactive substance has a much higher radioactivity concentration than the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77. That is, the cement solidified in the cement solidifying device 18 has a much lower sodium sulfate concentration than the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77. Therefore, it is possible to manufacture a cement solid without forming ettringite and dispose of the cement by burying.
[0034]
Next, the operation of the radioactive ion exchange resin treatment system to which the radioactive waste liquid separation device according to the embodiment of the present invention configured as described above is added will be described with reference to the flowchart of FIG.
[0035]
The concentrated waste liquid tank 77 stores radioactive substances such as Co (cobalt), Cs (cesium), Sr (strontium), and Nb (niobium), and A-mode concentrated waste liquid containing sodium sulfate. The A-mode concentrated waste liquid is separated into a supernatant liquid and a coagulated sediment in the concentrated waste liquid tank 77, and most of the sodium sulfate is dissolved and contained in the supernatant liquid. On the other hand, most of the coagulated precipitate is formed by a radioactive substance that becomes a hydroxide.
[0036]
Using the radioactive waste liquid separator according to the embodiment of the present invention, the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77 has a significantly lower sodium sulfate concentration and a higher radioactive concentration than the A-mode concentrated waste liquid. In order to separate the high-activity mixture and the low-activity concentration having a significantly lower activity concentration than the A-mode concentrated waste liquid, the mixture is first stored in the concentrated waste liquid tank 77 by the supernatant extraction pump 10. The supernatant liquid of the A-mode concentrated waste liquid is extracted and sent to the inorganic ion adsorbent tower 12 from below to above (S1).
[0037]
The supernatant sent to the inorganic ion adsorbent tower 12 contains a radioactive substance. This radioactive substance is adsorbed by the adsorbent provided inside the inorganic ion adsorbent tower 12 when the supernatant liquid is sent upward from below inside the inorganic ion adsorbent tower 12 (S2). .
[0038]
Therefore, from the upper part of the inorganic ion adsorbent tower 12, the radioactive substance is removed by the adsorbent, the supernatant liquid having a reduced radioactivity concentration is discharged, discharged to the cement solidifying device 16, and solidified therein. (S3).
[0039]
The supernatant liquid discharged to the cement solidifying device 16 has the same sodium sulfate concentration but lower radioactivity concentration than the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77. Such low-radioactive waste liquid can be disposed of in a simpler disposal facility than a solidified cement obtained by solidifying the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77 as it is. Therefore, by generating such a low-radioactive waste liquid from the A-mode concentrated waste liquid, it is possible to reduce the cost of landfill disposal.
[0040]
When the supernatant is thus cemented, the pure water supply system 13 supplies pure water in an amount corresponding to the extracted supernatant to the concentrated waste liquid tank 77 (S4). Thus, the coagulated sediment remaining in the concentrated waste liquid tank 77 after the supernatant is extracted is diluted. The coagulated sediment contains a part of sodium sulfate. Thus, by diluting the coagulated sediment, sodium sulfate is also diluted.
[0041]
The coagulated sediment diluted in step S4 is transferred from the concentrated waste liquid tank 77 to the batch tank 14. Then, the coagulated sediment is stored until the cement is solidified by the cement solidifying device 18 (S5).
[0042]
The diluted coagulated sediment stored in the batch tank 14 in step S7 is discharged to the cement solidifying device 18 when the cement solidifying device 18 is ready to solidify the cement. The adsorbent that has absorbed the radioactive substance in the inorganic ion adsorbent tower 12 is also discharged to the cement solidifying device 18. Then, these are mixed and cement-solidified in the cement-solidification device 18 (S6).
[0043]
The diluted coagulated sediment discharged to the cement solidifying device 16 has a much lower sodium sulfate concentration than the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77. On the other hand, the adsorbent discharged from the inorganic ion adsorbent tower 12 has a radioactivity concentration much higher than that of the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77 because it absorbs radioactive substances. .
[0044]
Therefore, the highly radioactive mixture obtained by mixing the diluted coagulated sediment discharged from the batch tank 14 and the adsorbent discharged from the inorganic ion adsorbent tower 12 is more effective than the A-mode concentrated waste liquid. Since the concentration of sodium sulfate is remarkably low, it is possible to manufacture a cement solid without forming ettringite. Thereby, it can be buried and disposed as a stable cement solidified body.
[0045]
As described above, in the radioactive ion exchange resin treatment system to which the radioactive waste liquid separation device according to the embodiment of the present invention has been added, the A-mode concentrated waste liquid that is the concentrated radioactive waste liquid is reduced by the above-described operation. It is possible to separate the radioactive waste liquid and the high radioactive mixture obtained by mixing the waste of the adsorbent and the diluted coagulated sediment, solidify each separately, and store them separately.
[0046]
The low radioactive waste liquid has the same sodium sulfate concentration and a low radioactivity concentration as compared with the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77. Accordingly, since the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77 can be disposed of in a simple landfill disposal facility as compared with a cement-solidified product obtained by solidifying cement, the cost of landfill disposal can be reduced.
[0047]
On the other hand, the high-activity mixture has a higher radioactivity concentration but a lower sodium sulfate concentration than the A-mode concentrated waste liquid stored in the concentrated waste liquid tank 77. As described above, since the concentration of sodium sulfate is low, it is possible to manufacture a cement solid that does not form ettringite. Thereby, it can be buried and disposed as a stable cement solidified body.
[0048]
Although the preferred embodiments of the present invention have been described with reference to the accompanying drawings, the present invention is not limited to such configurations. Within the scope of the invented technical concept of the claims, those skilled in the art will be able to conceive various changes and modifications, and those changes and modifications will be described in the technical scope of the present invention. It is understood that it belongs to.
[0049]
【The invention's effect】
As described above, according to the present invention, the A-mode waste liquid, the low-activity waste liquid having a lower radioactivity concentration than the A-mode waste liquid, and the sodium sulfate concentration significantly lower than the A-mode waste liquid and the higher radioactivity concentration It can be separated into highly radioactive mixtures and solidified for each.
[0050]
Furthermore, by storing and solidifying each solidified substance in another buried disposal facility in accordance with the level of radioactivity contained, ettringite can be suppressed and the solidified substance can be stabilized. A radioactive waste liquid separation device and a radioactive ion exchange resin treatment system including the same can be realized.
[Brief description of the drawings]
FIG. 1 is a system configuration diagram showing an example of a radioactive ion exchange resin treatment system to which a radioactive waste liquid separation device according to an embodiment of the present invention is added. FIG. 2 is a radioactivity to which a radioactive waste liquid separation device according to the embodiment is added. Flow chart showing the operation of the ion exchange resin treatment system. FIG. 3 is a system configuration diagram of a conventional radioactive ion exchange resin treatment apparatus. FIG. 4 is a conceptual diagram for explaining the principle of acid recovery in a sulfuric acid recovery unit. Of the time course of the radioactivity concentration in the sulfuric acid aqueous solution [Explanation of the symbols]
DESCRIPTION OF SYMBOLS 10 ... Supernatant extraction pump 12 ... Inorganic ion adsorbent tower 13 ... Pure water supply system 14 ... Batch tanks 16 and 18 ... Cement solidifier 62 ... Spent resin storage tank 63 ... Waste resin supply tank 64 ... Elutator 67 ... Elutation Used resin transport container 68 incinerator 69 sulfuric acid recovery unit 70 left chamber 71 diffusion dialysis membrane 72 right chamber 73 neutralizing tank 75 neutralizing agent tank 76 evaporator 77 concentrated waste liquid tank 78 liquid disposal Material processing system

Claims (2)

A waste liquid tank for storing a radioactive waste liquid containing sodium sulfate (Na ₂ SO ₄ );
The supernatant of the radioactive waste liquid stored in the waste liquid tank is supplied, comprising an adsorbent for adsorbing the radioactive substance contained in the supernatant, a radioactive substance removing means for generating a low radioactive liquid waste,
First solidifying means for solidifying the low radioactive liquid waste generated by the radioactive substance removing means,
A radioactive waste liquid separating apparatus comprising: a second solidifying means for mixing and solidifying a precipitate collected from the waste liquid tank and an adsorbent to which a radioactive substance has been adsorbed in the radioactive substance removing means. at least,
An eluent for passing an aqueous sulfuric acid solution for elution through the ion exchange resin to which the radioactive ion species are adsorbed, and eluting the radioactive ion species adsorbed on the ion exchange resin into the aqueous sulfuric acid solution for elution;
By adding sodium hydroxide (NaOH) as a neutralizing agent to an eluted sulfuric acid aqueous solution obtained by eluting a radioactive ion species into the above-mentioned aqueous sulfuric acid solution for elution, sodium sulfate (Na ₂ SO ₄ ) as a neutralized form is contained. A neutralization means for generating a neutralized waste liquid to be treated, and a radioactive ion exchange resin treatment system including a concentration means for concentrating the neutralized waste liquid to generate a concentrated neutralized waste liquid,
A waste liquid tank for storing the concentrated neutralized waste liquid generated by the concentration means,
An extraction pump for extracting a supernatant of the concentrated neutralized waste liquid stored in the waste liquid tank,
An adsorbent is provided inside, and the radioactive substance contained in the supernatant liquid extracted by the extraction pump is adsorbed on the adsorbent to remove the radioactive substance from the supernatant, thereby concentrating and neutralizing. A radioactive substance removing means for producing a low radioactive waste liquid having a lower radioactive concentration than the waste liquid,
First solidifying means for solidifying the low radioactive liquid waste generated by the radioactive substance removing means,
Sediment collecting means for separating and collecting the sediment by adding water to the sediment of the concentrated neutralized waste liquid stored in the waste liquid tank,
A high radioactivity mixture having a higher radioactivity concentration than the concentrated neutralized waste liquid obtained by mixing the precipitate collected by the sediment collection means with an adsorbent to which a radioactive substance has been adsorbed in the radioactive substance removing means. A system for treating a radioactive ion exchange resin, comprising a radioactive waste liquid separation device comprising:

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