JPH0727076B2 - Treatment method and equipment for radioactively used ion exchange resin - Google Patents

Description

TECHNICAL FIELD The present invention relates to a method of treating radioactive used ion-exchange resin (hereinafter referred to as waste resin) generated from a nuclear power plant and a treatment facility, and particularly to a method of treating a radioactive level. The present invention relates to a treatment method and treatment equipment suitable for treating a relatively high waste resin.

[Conventional technology]

Waste resin generated from a nuclear power plant can be roughly classified into low-level waste resin having a relatively low radioactivity level and high-level waste resin having a relatively high radioactivity level. Here, the high level waste resin refers to a waste resin having a radioactivity level of approximately 1 to 10 μCi / g, and the low level waste resin has a radioactivity level of approximately 0.
01 to 1 μCi / g of waste resin. Taking a BWR power plant as an example, what is generated from the condensate purification system (condensate demineralizer, condensate demineralizer, etc.) is equivalent to low-level waste resin, and what is generated from the reactor water purification system is Corresponds to high-level waste resin.

Among these, various treatment methods (plastic solidification method, thermal decomposition method, etc.) have been developed for low-level waste resin, and the final solidified product obtained will be finally disposed of on land or the like in the near future. is there.

On the other hand, the high-level waste resin is currently stored in the tank in the power plant and is not subjected to solidification treatment. There are two reasons for this. First, the high-level waste resin has a radioactivity level 10 to 100 times higher than that of the low-level waste resin, and the conventional solidification treatment method poses a problem such as exposure of workers. Secondly, even if the solidification treatment is carried out, the final solidified body has a high radioactivity level and thus cannot be finally disposed of. Since the final solidified body still has to be stored in the power plant, the merit of the solidification treatment is small. is there.

Therefore, regarding handling of high-level waste resin, JP-A-61
The method shown in −17995 has been tried. That is, this is a method in which a part of the radionuclide contained in the waste resin is extracted with a tartaric acid solution to reduce the radioactivity level of the waste resin and then the waste resin is incinerated to reduce its volume. According to this method, the radioactivity level of the waste resin is reduced by 10 to 50%, so that it is possible to surely reduce the exposure of workers during the treatment of the waste resin. However, compared to so-called low-level waste resin, the level of radioactivity is still high, so the problem of worker exposure remains. Furthermore, the incinerated ash after the treatment by incineration can be solidified with plastic or the like to give a final solidified body, but since this final solidified body still has a considerably high radioactivity level, it can be finally disposed of. No, so the final solidified material still has to be stored in the power plant.

The above describes the conventional technology regarding the treatment of high-level waste resin, but next, regarding the treatment of conventional low-level waste resin, all low-level waste resin generated from nuclear power plants is treated as radioactive waste and All of them are stored in the nuclear power plant until the final disposal, and there is a problem that they cannot be discarded as general waste and the amount of storage is enormous.

[Problems to be solved by the invention]

As described above, the conventional waste resin treatment technology cannot sufficiently increase the separation efficiency of radioactivity in the high-level waste resin treatment technology, and the final disposal of the final solidified product (especially on land There is a problem in terms of compatibility with (disposal) and reduction of exposure of workers at the time of processing, and in the processing technology of low-level waste resin, all of this must be treated as radioactive waste and final disposal. Although it is possible to prepare a solidified material that can be manufactured, there is a problem in that the storage amount becomes enormous.

In view of the above points, the object of the present invention is to provide a high-level waste resin with a final solidified product having excellent compatibility with final disposal and also reduce the exposure dose of workers, and with respect to a low-level waste resin. The object of the present invention is to provide a method and equipment for treating radioactive waste resin capable of significantly reducing the amount of final disposal as radioactive waste and reducing the storage amount.

[Means for solving problems]

In order to achieve the above object, the present invention as a processing method,
A radioactive used ion exchange resin generated in a nuclear power plant and a neutral salt solution or an acidic solution from a solution supply tank are supplied to a radioactive substance separation tank, and the resin in the solution in the separation tank is mechanically treated. The radioactive clad is physically separated from the resin by applying a force, and at the same time, the radioactive ions are chemically separated by an ion exchange reaction to recover the resin as a resin having a lower activity level, while the radioactive clad is separated. And the above-mentioned solution to which the radioactive ions have been transferred is passed through the clad recovery device and the ion recovery device in that order as a waste liquid, and the radioactive clad and the radioactive ions are recovered in each, and the waste liquid after the recovery is circulated to the solution supply tank. It is characterized by being reused.

Further, similarly, the present invention as a treatment facility is a radioactive substance separation tank to which both a radioactive spent ion exchange resin generated in a nuclear power plant and a neutral salt solution or an acidic solution from a solution supply tank are supplied, By physically applying a mechanical force to the resin in the solution in the tank, the radioactive clad is physically separated from the resin, and at the same time, radioactive ions are chemically separated by an ion exchange reaction. Separation means in combination with separation means, resin recovery means for recovering low radioactivity level resin after radioactive material is separated, the radioactive clad and the solution to which radioactive ions have transferred are used as a waste liquid, and the radioactive clad is recovered from the waste liquid. Clad recovery means, and ion recovery for recovering radioactive ions from the waste liquid after clad recovery, which is downstream of the clad recovery means It is characterized in that the effluent from stage and the ion collection means with a solution recycling means for circulating the said solution supply tank.

[Action]

The high-level waste resin generated from the reactor water purification system of BWR power plants usually contains radioactive substances in the range of 1 to 10 μCi / g, and its radioactive nuclide is mainly ⁶⁰ Co. ¹³⁷ Cs etc. are also included. On the other hand, in the conventional technologies such as the plastic solidification method and the incineration method that have been conventionally used for low-level waste resin, the final solidified body (considering a 200-liter drum) can be filled with about 50 to 1000 kg of waste resin. If this conventional technology is directly applied to high-level resin, 1-10
μCi / g × 50 to 1000 kg / drum = 0.05 to 10 Ci / drum, that is, 0.05 l to 10 l of radionuclide is the final solidified product 200 l
It will be included in the drum. At this time, the surface dose rate of the final solidified body is about 400 mR / h to 80 R / h.

Although the disposal conditions for the final solidified land disposal are currently undecided, it is considered that the surface dose rate should be less than 200mR / h. Therefore, there is a problem that the final solidified product cannot be finally disposed of by treating the high-level waste resin with the conventional technique.

Therefore, the present inventors considered that a part of the radionuclide contained in the high level waste resin should be separated to lower the level of the waste resin to enable land disposal. In this case, the disposal condition is 200 mR / h, and the surface dose rate is 400 m when the high-level waste resin is made into the final solidified body by the conventional method.
Considering that it is R / h-80R / h, it is considered that at least 50% (200mR / h ÷ 400mR / h = 1/2) is required for the separation efficiency of nuclides, and if a further safety factor is considered. It is clear that higher separation efficiency (99.75%) is desirable.

Based on the above considerations, in order to find a method for efficiently separating the radionuclide from the waste resin, first, we investigated how the radionuclide was adsorbed to the waste resin. I understand.

Most of the radionuclides adsorbed on the waste resin are corrosion products such as ⁶⁰ Co, and in some cases, even in small amounts.
Fission products such as ¹³⁷ Cs are also included. Below, ⁶⁰ Co is representative of corrosion products and ¹³⁷ Cs is representative of fission products.
Will be explained. Most of ⁶⁰ Co forms a complex with oxides (non-radioactive Fe ₂ O ₃ etc.) in reactor water,
It is physically adsorbed on the waste resin in the form of so-called clad called ferrite. On the other hand, it was newly found that ¹³⁷ Cs and some ⁶⁰ Co are adsorbed on the waste resin by ion exchange reaction. From the above findings, it was found that both the physically adsorbed clad and the ion adsorbed ion must be removed in order to efficiently separate the radionuclide in the waste resin.

Next, when I examined how to remove these,
Regarding the separation of the physically adsorbed clad, it was found that the waste resin can be removed at a high removal rate of 90 to 99.9% by immersing the waste resin in the liquid and applying mechanical force (stirring or ultrasonic wave). Regarding separation of radionuclides adsorbing ions, the radionuclides adsorbing ions are ¹³⁷ C.
In addition to s, there are ⁶⁰ Co, both of which are cationic,
Waste resin is treated with neutral salt solution such as NaNO ₃ , NaCl, Na ₂ SO ₄ or HNO
If it is immersed in an acidic solution such as ₃ , HCl, H ₂ SO _4, most of the radionuclides (efficiency 90% or more) will undergo normal ion exchange reactions.
Was found to migrate into the solution, allowing the separation of radionuclides.

The invention thus removes from the waste resin the radioactive substances which are the main cause of its radioactivity.

Waste wood at a lower level than the original waste resin obtained by separating and removing radioactive materials in this manner is then
A final solidified product that meets the final disposal conditions can be obtained by performing an appropriate volume reduction and solidification process. This allows the final solidified body to be finally disposed of, eliminating the need for long-term storage within the power plant.

In addition, by separating the waste resin at a lower level and the waste material at a higher level than the original waste resin, it is possible to take measures such as shielding according to the radioactivity level, so that the exposure dose of workers can be reduced. .

The above describes the treatment of high-level waste resin, but if this principle is applied to low-level waste resin, most of the low-level waste resin should be treated as general waste and have extremely low levels of radioactivity. Therefore, only a small amount of high-level waste containing the separated radionuclide should be made into the final solidified material suitable for final disposal.

〔Example〕

One embodiment of means for separating the clad adsorbed on the waste resin will be described with reference to FIG. The waste resin 2 adsorbing the clad 1 is continuously supplied to the clad separation tank 3, and the liquid 4 is continuously supplied to the clad separation tank 3 from above, and a part of the liquid 4 is provided in the clad separation tank 3. It is continuously discharged through the perforated plate 5 thus formed. At this time, the ultrasonic wave 6 provided in the clad separation tank 3 causes the liquid 4 and the waste resin 2 and the clad 1 to undergo mechanical vibration. As a result, the clad 1 physically adsorbed on the waste resin 2 is peeled off,
The liquid 4 is discharged to the outside of the clad separation tank 3. By applying the mechanical peeling force in this way, ⁶⁰ Co which forms the clad and a part of the clad physically adsorbed by the waste resin
Confirmed that most of them (removal rate 90 to 99.9%) can be separated.

Although FIG. 2 shows the case where processing is performed continuously, patch processing is also possible. That is, when the waste resin 2 adsorbing the clad 1 is immersed in the liquid 4 and mechanical vibration is applied, the particle size of the waste resin 2 is as large as 30 to 500 μm, whereas the particle size of the clad 1 is 0.1 to Since it is as small as 10 μm, most of it moves into the liquid 4 and becomes a colloidal state. That is, the clad 1 can be separated by a patch process.

As means for applying mechanical vibration, in addition to the ultrasonic vibration, mechanical stirring with a stirring blade or a combination of both is also possible.

As the liquid 4, a neutral salt solution or an acidic solution is used, and at the same time as the physical separation of the clad adsorbed on the waste resin, the radionuclide adsorbed as an ion on the waste resin can be separated.

Next, an embodiment of separating the radionuclide from the waste resin of the present invention will be described with reference to FIG. In Figure 3, ⁶⁰ Co, ¹³⁷ C
The waste resin 2 adsorbing s or the like in the form of clad or ions is supplied from the tank 2 to the clad separation tank 3. A neutral salt solution or an acidic solution 7 is supplied to the clad separation tank 3, and mechanical vibration is applied by the ultrasonic oscillator 6. As a result, the physically adsorbed clad is separated from the waste resin 2 by mechanical peeling, and the ion adsorbed ion is separated from the waste resin 2 and transferred to a neutral salt solution or an acidic solution. As a result, the radioactive level of the waste resin 2 is lowered. This low-level waste resin can be applied with the plastic solidification and incineration methods already developed for ordinary low-level waste resin, and the final solidified product can be disposed of on land in many cases.

On the other hand, most of the radionuclides adsorbed on the waste resin 2 are
The liquid is transferred to a solution and becomes a waste liquid 8. The radioactive clad 1 in the waste liquid is a clad recovery device 9 using a filter such as a hollow fiber membrane filter, an electromagnetic filter or a mechanical means such as a gravity sedimentation method. Radioactive ions can be collected and concentrated by the ion recovery device 10 using an adsorbent such as zeolite or chelate resin. That is, as shown in FIG. 3, the radionuclide in the waste liquid 8 is concentrated by the clad recovery device 9 and the ion recovery device 10. Secondary waste with a high level of radioactivity is generated from this, but it is often difficult to dispose of it on land, so store it in the power station. However, this secondary waste amount is extremely small compared to the original amount of waste resin, so it does not cause a big problem even if it is stored in the power plant.

The efficiency of separating radionuclides from waste resin 2 is 50 to 99.75%
That is all. The reason for this is that the surface dose rate of the final solidified product is 400 mR / h when the waste resin is solidified as it is as described above.
This is because the surface dose rate must be less than 200 mR / h to dispose of the final solidified material on land.

Next, specific examples of the present invention will be described below.

Example 1 In this example, waste resin generated from a reactor water purification system of a nuclear power plant is treated, and FIG. 1 shows a system diagram of the treatment system used in this example.

As a result of a preliminary analysis of the radionuclide adsorbed on the waste resin, the following was found. The major nuclides were ⁶⁰ Co and ¹³⁷ Cs, and the specific activities were ⁶⁰ μCi / g for ⁶⁰ Co and 0.5 μCi / g for ¹³⁷ Cs. Further, about 95% of ⁶⁰ Co was physically adsorbed on the waste resin together with the clad, and about 5% of ⁶⁰ Co and about 100% of ¹³⁷ Cs were ion adsorbed on the resin by ion exchange.

After such waste resin 2 is supplied to the clad separation tank 3, about 10% NaNO ₃ solution 12 is continuously supplied from the neutral salt supply tank 11 and mechanical vibration is generated by the ultrasonic oscillator 6. added. As a result, the clad 1 containing ⁶⁰ Co and the ion-adsorbed ¹³⁷ Cs and ⁶⁰ Co both migrated into the NaNO ₃ solution and flowed out to the waste liquid 8 side. After that, a clad settling tank 13 was provided, and the clad 1 in the NaNO ₃ solution was separated by the static method. Waste liquid after clad separation 8
Further, it was led to the ion adsorption tower 14. Zeolite 15 and chelate resin 16 were filled therein as an adsorbent, and ionic ¹³⁷ Cs was removed by zeolite 15 and ionic ⁶⁰ Co was removed by chelate resin 16. The NaNO ₃ solution from which the radionuclide was removed by such a method was returned to the neutral salt supply tank 11 and reused.

As a result, 99% of the radionuclide adsorbed on the waste resin was separated, the clad settled in the clad settling tank 13, and the ions transferred to the adsorbent in the ion adsorption tower 14. Further, the waste resin having a specific activity reduced to 0.1 μCi / g (corresponding to a separation efficiency of about 99%) was solidified by the plastic solidification device 17 to be the final solidified body 18. This final solidified body 18 is 200 l
It was in the form of a drum, and about 100 kg of waste resin was filled in each drum. At this time, the radioactivity content of the final solidified product was 10 mCi / drum, and the surface dose rate was 70 mR.
It was / h. Therefore, this final solidified material can be subjected to final disposal such as land disposal, like conventional low-level waste. In addition, the clad 1 settled in the clad settling tank 13 and the zeolite 15 after adsorbing ionic ⁶⁰ Co and ¹³⁷ Cs
Since the specific activity of the chelate resin 16 is as high as about 100 μCi / g, the chelate resin 16 was solidified with cement, which is an inorganic substance, to obtain the high-level solidified body 19. Since this high-level solidified material 19 does not meet the final disposal conditions, it was necessary to store it in the power plant.

When 1000 kg of waste resin was treated by the above method, the radioactive nuclides in the waste resin could be separated with an efficiency of 99%, 10 final solidified bodies 18 that could be finally disposed of, and high level solidification that needs to be stored in the power plant One body 19 occurred. On the other hand, when the waste resin was directly plasticized without such treatment, 10 high-level solidified materials that had to be stored in the power plant were generated. That is, according to the present embodiment, the number of high-level solidified bodies that need to be stored in the power plant is about 1 /
It can be reduced to 10, and the rest can be finally disposed of.

In this embodiment, since the principle is extremely simple, automation is easy,
Moreover, since the high radioactivity level is limited to the clad settling tank 13, the ion adsorption tower 14, etc., it is easy to intensively enhance the shielding. Therefore, it is possible to easily take measures to reduce the exposure of workers.

Further, in the present example, the waste resin having a reduced specific activity was plasticized, but the known low-level waste resin treatment technology, such as asphalt solidification, cement solidification, cement vitrification, incineration or solidification after pyrolysis, is used. It goes without saying that both can be used, and also that these treatment devices can be shared with those already installed in the power plant for low-level waste.

In addition, in order to create the high-level solidified body 19, it goes without saying that glass solidification, rock solidification, hydrothermal solidification, cement vitrification, and the like can be used in addition to the above-mentioned cement solidification.

As the adsorbent filled in the ion adsorption tower 14, titanium adsorbent, mordenite, etc. can be used in addition to zeolite and chelate resin.

Although NaNO ₃ was used for ion separation in this example, other neutral salt solutions such as Na ₂ SO ₄ and NaCl can also be used.

Example 2 The main radionuclides adsorbed on the waste resin to be treated in this Example 2 are ⁶⁰ Co and ¹³⁷ Cs, which is the same as in Example 1, but ⁶⁰ Co is wasted almost 100% together with the clad. This is different from Example 1 in that it is physically adsorbed to the resin, and almost 100% of ¹³⁷ Cs is ion-adsorbed to the waste resin. FIG. 4 shows a systematic diagram of the processing system used in the second embodiment.

After the waste resin 2 is supplied to the clad separation tank 3, about 10% Na ₂ SO ₄ solution 20 is continuously supplied from the neutral salt supply tank 11 and the stirring blade 21 provided in the clad separation tank 3 is used. Mechanical stirring power was applied. As a result, both the clad 1 containing ⁶⁰ Co and the ion-adsorbed ¹³⁷ Cs migrated into the Na ₂ SO ₄ solution and flowed out to the waste liquid 8 side. After that, a hollow fiber membrane filter 22 for separating the clad 1 is provided, and further, after that, an ion adsorption tower 14 for removing ¹³⁷ Cs is provided. In the ion adsorption tower 14, Nu
The copper-impregnated zeolite 23 shown in cl.Technol. Vol 58, pp.242-247 (1982) was packed, and after efficiently removing ¹³⁷ Cs, the waste liquid 8 was reused as in Example 1.

As a result, 98% of radionuclides adsorbed on waste resin
Was removed, and the low-level waste resin was made into the final solidified body 18 by the cement vitrification device 24. This final solidified body
No. 18 had a low surface dose rate of 50 mR / h and could be finally disposed of.

On the other hand, the clad separated by the hollow fiber membrane filter 22 is
A cement vitrification device 24 'similar to the cement vitrification device 24 was used to obtain a high-level cement vitrification product 25, which had a high surface dose rate of 10 R / h and could not be finally disposed of. However, since the major radionuclide contained in this high-level cement vitrified material 25 is ⁶⁰ Co, which has a short half-life of 5 or 6 years, this high-level cement vitrified material
25 can also be finally disposed of if stored in the power station for decades and waiting for radioactivity to decay.

On the other hand, ¹³⁷ Cs separated by the ion adsorption tower 14 and having a long half-life of 30.2 years was hydrothermally solidified together with the used adsorbent to become a high-level rock solidified body 26. This high level rock solidified body 26
Has a high surface dose rate of 20 R / h, and it can hardly be expected to attenuate radioactivity, so it will be necessary to store it in the power plant for the time being and make final disposal suitable for disposal of high-level waste such as underground disposal in the future. Is.

When 1000 kg of waste resin was processed according to this example, about 10 final solidified bodies 18 which can be finally disposed (200 l drum can conversion,
The same shall apply hereinafter), 0.8 high-level cement vitrified solids 25 that could be finally disposed of in the future, and 0.1 high-level rock solidified solids 26 that were difficult to finalize for the time being were generated. Thus, by classifying the final solidified matter based on the radioactivity level and the half-life of the nuclide, the amount of waste that needs to be stored in the power plant can be further reduced.

In addition, in the present Example 2 and Example 1, as a final solidified body
Although we considered a 200-liter drum, it is of course possible to use a rectangular container, a highly sound container, etc. Depending on the container, it may be processed so that the final disposal conditions are satisfied.

Example 3 The present Example 3 is to improve the separation efficiency of the clad by treating the resin with an acidic solution instead of treating the resin with the neutral salt solution in Example 1 or 2. Is.

Normally, the clad is only physically adsorbed to the waste resin, but if the waste resin is stored in the tank for a long time, the clad may stick to the waste resin. In addition, the clad separation efficiency is low only by applying mechanical force. FIG. 5 shows the separation efficiency as a function of the separation time (time when ultrasonic vibration is applied) when the clad adhered to the waste resin is separated by ultrasonic vibration. It can be seen that the clad separation efficiency is higher when the pH is adjusted to ₃ by adding a small amount of HNO ₃ to NaNO ₃ as compared with the case where the NaNO ₃ solution of pH 7 is used. Further, FIG. 6 shows the separation efficiency as a function of the pH of the solution when the separation time is 20 minutes. From this, it can be seen that by setting the pH of the solution to 4 or less, high clad separation efficiency can be obtained even when the clad is fixed to the resin. The reason why such a high separation efficiency is obtained is that a part of the clad that has been fixed is chemically eluted with an acid. On the other hand, the pH of the solution
When the value is 1 or less, the ion adsorption efficiency by the zeolite, chelate resin, etc. in the ion adsorption tower 14 decreases. Therefore, it is desirable that the pH of the solution be in the range of 1 to 4 for the fixed clad.

When the solution is reused and repeatedly used as in Examples 1 and 2, the pH of the solution may change in the long term. In this case, a suitable place of the solution circulation line (for example, between the clad sedimentation tank 13 and the ion adsorption tower 14 in FIG. 1)
In addition, a pH adjusting tank may be provided.

Example 4 In each of the above Examples, the high level waste resin having a relatively high radioactivity level generated from the reactor cleaning system of a nuclear power plant is targeted, whereas the Example 4 is The present invention is applied to a low-level waste resin having a relatively low radioactivity level generated from a condensate purification system, a waste treatment system and the like. In this case, a new effect is created. That is, at present, if the radionuclide is separated from the low level resin generated from the nuclear power plant by the same method as in the above example, most of the low level waste resin has extremely low specific activity to the extent that it can be regarded as general waste. And its treatment becomes easy. At the same time, a small amount of waste having a high radioactivity level generated from the waste resin is generated, but this may be finally disposed of by land disposal or the like after solidification treatment.

As described above, when the present invention is applied to the low-level resin, the amount of waste required for final disposal is significantly reduced, and the cost can be reduced. The efficiency required for clad separation and ion separation will be determined if the level that can be regarded as general waste is determined.
It can be determined from the value and the specific activity of the resin.

〔The invention's effect〕

According to the present invention, the radionuclide can be separated from the waste resin with high separation efficiency by simultaneously using the physical separation means and the chemical separation means at the same time. Most of the ion exchange resin (waste resin) can be finally disposed of. Further, the neutral salt solution or acidic solution used as the chemical separation means,
The radioactive clad and radioactive ions migrate from the waste resin and become waste liquid, but after the migrated radioactive substance is collected,
Since it is recycled and reused, the amount of waste liquid to be finally disposed of can be reduced by itself. Furthermore, since the work range for handling high-level waste in the treatment facility is limited, the worker's exposure can be reduced, which is effective in reducing the exposure in combination with facilitating automation of the treatment facility. Furthermore, when applied to the treatment of low-level waste resin, the amount of radioactive waste required for final disposal can be greatly reduced.

[Brief description of drawings]

FIG. 1 is a system diagram of a waste resin treatment facility according to a first embodiment of the present invention, FIG. 2 is a schematic diagram showing an embodiment of a clad separating means, and FIG. 3 is a system diagram of a principle embodiment of the present invention. FIG. 4 is a system diagram of a waste resin treatment facility according to Example 2, FIG. 5 is a diagram showing a relationship between clad separation efficiency and separation time, and FIG. 6 is a solution.
It is a figure which shows the relationship between pH and clad separation efficiency. DESCRIPTION OF SYMBOLS 1 ... Clad, 2 ... Resin 3 ... Clad separation tank, 6 ... Ultrasonic oscillator 9 ... Clad recovery device, 10 ... Ion recovery device 11 ... Neutral salt solution supply tank 13 ... Clad sedimentation tank, 14 ... Ion adsorption tower 18 … Final solidified product, 19… High level solidified product 21… Stirrer, 22… Hollow fiber membrane filter 25, 26… Final solidified product

 ─────────────────────────────────────────────────── ─── Continuation of front page (72) Fumio Kawamura 1168 Moriyama-cho, Hitachi-shi, Ibaraki Energy Research Institute, Nitrate Manufacturing Co., Ltd. (72) Inventor Shin Tamada 3-1-1, Saiwai-cho, Hitachi, Ibaraki Hitachi, Ltd. Hitachi factory (56) Reference JP-A-57-44899 (JP, A) JP-A-61-254900 (JP, A) JP-A-60-166042 (JP, A)

Claims (6)

[Claims] 1. A radioactive spent ion exchange resin generated in a nuclear power plant and a neutral salt solution or an acidic solution from a solution supply tank are supplied to a radioactive substance separation tank, and the above-mentioned solution is used in the solution in the separation tank. The radioactive clad is physically separated from the resin by applying a mechanical force to the resin, and at the same time, the radioactive ions are chemically separated by an ion exchange reaction to recover the resin as a resin having a lower radioactivity level. On the other hand, the radioactive clad and the solution to which the radioactive ions are transferred are passed through the clad recovery device and the ion recovery device in that order as waste liquid, and the radioactive clad and the radioactive ion are recovered in the respective waste liquids, and the waste liquid after the recovery is supplied to the solution. A method for treating radioactively used ion-exchange resin, characterized in that it is circulated to a tank for reuse. 2. The radioactive substance according to claim 1, wherein the physical separation of the radioactive clad from the resin is performed by applying a mechanical force by at least ultrasonic vibration or stirring. Treatment method of used ion exchange resin. 3. In the case where the solution is an acidic solution,
The method for treating a radioactive used ion exchange resin according to claim 1, wherein the pH of the acidic solution is in the range of 1 to 4. 4. A radioactive substance separation tank to which both a radioactive used ion exchange resin generated in a nuclear power plant and a neutral salt solution or an acidic solution from a solution supply tank are supplied, and in the solution in the separation tank, Separation means using both physical separation means and chemical separation means for physically separating the radioactive clad from the resin by applying a mechanical force to the resin and at the same time chemically separating radioactive ions by an ion exchange reaction A resin recovery means for recovering a low radioactivity level resin after the radioactive substance is separated, a clad recovery means for recovering a radioactive clad from the waste solution by using the solution in which the radioactive clad and radioactive ions have been transferred as a waste solution, and the clad Ion recovery means downstream of the recovery means for recovering radioactive ions from the waste liquid after clad recovery, and the ion recovery means Treatment equipment of waste radioactive spent ion exchange resin comprising the solution recycling means for circulating the said solution supply tank. 5. The physical separating means includes ultrasonic vibrating means and / or stirring means for applying a mechanical force to the resin in the solution. Treatment equipment for radioactively used ion exchange resin according to the item. 6. In the case where the solution is an acidic solution,
The pH of the solution is in the range of 1 to 4, and the pH is set to 1 in the waste liquid flow path between the clad recovery means and the ion recovery means.
The treatment facility for radioactively used ion exchange resin according to claim 4, further comprising a pH adjusting tank for adjusting the pH to the range from 4 to 4.