JP3024416B2 - Radioactive waste treatment method - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

The present invention relates to a method for treating radioactive waste.
Relates to the law, in particular, nuclear facilities (power plants, reprocessing plants) relates to a process how suitable radioactive waste to handle radioactive waste from.

[0002]

2. Description of the Related Art From nuclear facilities, especially nuclear power plants, various radioactive wastes such as concentrated waste liquids and dried powders and pellets thereof, incinerated ash, incombustible miscellaneous solids, and used ion exchange resins are generated. These radioactive wastes are subjected to an appropriate volume reduction treatment, and then solidified stably in a dedicated container with a solidifying material such as cement or plastic. The solidified material thus produced is used to determine the radionuclide concentration, surface dose rate,
After undergoing rigorous inspections regarding strength, etc., only those that meet the burial storage standard values are buried and stored in storage facilities built underground.

[0003] In order to store the solidified material buried in the ground, an accurate radioactivity inventory per solidified material is evaluated.
It is necessary to check with the buried standard value. Conventionally, an unloading management system that measures the amount of radioactivity per solid after solidification has already been developed. Specifically, those described in Japanese Patent Application Nos. 61-26345 and 2-157340 have been developed. Yes. In radioactivity measurement, since a radiation detector is installed outside the solidified body for measurement, γ-rays with strong penetrating power can be measured, but α- and β-rays with weak penetrating power are shielded by the solidified body and measured. Can not. Therefore, transuranium elements (α nuclides) that mainly emit α rays
And nuclides that emit only β rays (β nuclides: Ni-63, C-
14) is difficult to measure directly.

[0004] Therefore, for such nuclides, the coefficients (scaling factors) statistically obtained from the correlation with Co-60 or Cs-137 that emit γ-rays are used. Inferred from measurements. In actual operation, the safety side is evaluated by applying a safety factor that covers statistical variations.

[0005]

The conventional radioactive waste treatment system has the following two problems. For one thing, since radiation measurement is performed after the solidified body is created, if the solidified body exceeding the burial standard value occurs, continue the site storage again or crush the nonconforming solidified body again and dilute and solidify it It is necessary to take additional measures such as. Second, when the scaling factor method is applied to α and β nuclides, it is evaluated unnecessarily on the safe side, and it may be judged that it is not suitable even though it is actually a solid that can be carried out That is. In particular, when reducing (concentrating) radioactive waste such as melting and compression volume reduction, the tendency is strong. The purpose of the present invention is to reduce the surface dose rate.
Prevent the generation of solidified waste exceeding the management standard value
And a method for treating radioactive waste.

[0006]

A feature of the present invention that achieves the above object is to measure the radioactivity information of radioactive waste,
Based on the radioactivity information, determine the disposal method for the radioactive waste.
Radioactive waste based on this determined treatment method.
A method of treating radioactive waste for treating waste, comprising:
The determined processing method is that the radioactivity information is lower than a reference value.
The radioactive waste determined to be high and the radioactivity information
For mixing with radioactive waste lower than the reference value
It is to be. Preferably, among the above-mentioned problems, in order to solve the first problem, before reducing the volume of radioactive waste, measure the amount or concentration of radioactive material contained in the waste, and do not exceed the control standard value. The amount of waste per solidified body is controlled. Before solving the second problem, a part of the waste is sampled, the concentrations of α-nuclide and β-nuclide are actually measured, and the amount of waste per solid is controlled so as not to exceed the control standard value .

The radioactive waste referred to herein is preferably
Nuclear facilities, especially nuclear power plants, include concentrated waste liquids and their dried powders and pellets, incinerated ash, incombustible miscellaneous solids, and radioactive waste such as used ion exchange resins. Also good
More preferably, the amount of waste is controlled by a waste weighing or metering device, which adjusts the amount of waste input per solid or the amount of radioactive waste to be reduced in volume. Or by mixing multiple wastes with different radionuclide concentrations.

[0008] In a process in which waste is formed into a single lump after processing such as pressing or melting, the radionuclide concentration and weight of the waste to be input before processing are measured, and the amount of waste per batch is determined. It is desirable to control the type.

[0009]

[Action] The radioactivity information is judged to be higher than the reference value.
Radioactive waste and radiation whose radioactivity information is lower than its reference value
Radioactive waste
And the generation of solidified waste exceeding the management standard value of the surface dose rate can be prevented. Preferably, as described above, before reducing the volume of the waste, the amount or concentration of the radioactive material contained in the waste is measured, and the amount of the waste per solidified body is controlled so as not to exceed the control standard value. By doing so, it is possible to prevent the generation of solids that do not conform to the control standard values,
Additional measures for incompatible solids can be eliminated.

[0010] Preferably, before reducing the volume of the waste, a part of the waste is sampled, the concentrations of α- and β-nuclides are actually measured, and the solidified body 1 is kept so as not to exceed the burial standard value.
By controlling the amount of waste per body, when the scaling factor method is applied, it is possible to prevent the solid body from being unnecessarily evaluated as being unsuitable because it is unnecessarily evaluated as being unsafe. .

[0011]

FIG. 1 shows an embodiment of the present invention. This embodiment relates to a system suitable for solidifying radioactive waste in a slurry state, such as used ion exchange resin (waste resin) and concentrated waste liquid generated from a nuclear facility, with cement.

The basic processing flow of this embodiment will be described with reference to FIG. Waste liquid or waste sludge is stored in a storage tank after being concentrated at an appropriate magnification. In treating such a waste liquid, the inside of the tank is uniformly mixed and stirred, and then a part is sampled to measure the concentration of the radionuclide. Based on the measured values, it is possible to evaluate the nuclide concentration in the case where the solid is finally solidified, and the following processing method is determined from the following by collating with the burial control standard value of the solidified solid. The treatment method is selected from the following methods in ascending order of the radioactivity level of the waste liquid.

(1) Further concentration (volume reduction) (2) No treatment (3) Dilution by mixing with waste liquid or another waste in another tank having a low radioactivity level (4) Return to tank for long-term storage (radiation) After the treatment, the waste liquid is dried and pulverized and then mixed with the solidifying material and solidified, or mixed with the solidifying material and solidified as it is. In the above, before volume reduction by dry powderization,
Although the radioactivity concentration and the like are measured, it can be measured before concentration.

Next, a system suitable for solidifying a used ion exchange resin (waste resin) with cement will be described with reference to FIG. The waste resin is a waste resin tank (A) 1 and a waste resin tank (B) according to the radionuclide concentration.
2 As an example, the nuclide (Co-60) concentration of the resin stored in the tank A is 10 times higher than the burying control standard value of the solidified body, and the nuclide concentration of the resin stored in the tank B is 1 standard value. / 100 will be described. After the resin slurry is drained by the dehydrator 3, the resin slurry is introduced into the mixer 4 in a set amount. A solidifying material tank 5 and an additive water tank 6 are connected to the mixer 4, and a valve and a feeder are installed so that each can be charged in a fixed amount. After these are sufficiently mixed by a mixer, they are poured into a solidification container 7, covered and cured to form a solid. After the solidified material is completely cured, the amount of radionuclides in the solidified body is checked by the inspection device 8, and only the solidified body that has cleared the burial control standard value is carried out.

In this example, the composition was set as shown in Table 1.

[0016]

[Table 1]

In this composition, the waste resin is diluted to 1/3 by weight. Therefore, when 100 kg of resin was supplied from the tank A to produce a solidified body without implementing the present invention, it was found that the inspection equipment exceeded the burial control standard value,
The solidified body cannot be carried out. Therefore, in the present invention, the sampling line 9 from each waste resin tank is installed, and the weight and radioactivity are measured by the measuring / measuring unit 10, and then the amount of waste resin to be charged is set. The radioactivity measurement is carried out by installing a NaI scintillator or a semiconductor detector configured using a standard radiation source in advance and connecting a multi-channel analyzer to obtain a counting rate for each nuclide. In this example, γ
The nuclides Co-60 and Cs-137 were measured, and the other α and β nuclides were evaluated by multiplying the previously described scaling factor by a safety factor of 10. In this example, A
The concentration of Co-60 in the resin stored in the tank B is 10 times higher than the buried management reference value of the solidified body, and the concentration of the resin stored in the tank B is 1/100 of the reference value. As I was able to recognize it before, I added 20 A resin to the tank.
80 kg of the resin in the kg and B tanks was weighed and charged into the mixer. In this case, the resin in the A tank is diluted 3 × 5 = 15 times, so that the solidified body can be carried out. Furthermore, using the nuclide concentration values measured by the measurement and measurement unit, calculate the surface dose rate of the solidified material after cement solidification, and evaluate the conformity of the surface dose rate with the burial control standard value and transport standard value. In addition, it is possible to control the amount of waste resin charged. In addition, by installing a β-ray detector such as a plastic scintillator in the measurement / measurement unit, Ni
It becomes possible to directly measure β nuclides such as −63, and it is not necessary to consider a safety factor as in the scaling factor method. In addition, as a method of controlling the mixing ratio between the solidified material and the waste using the measurement results, only 20 kg of the resin in the A tank is measured and charged into the mixer, and the charged amount of the solidified material and water is 170 kg.
It is also possible to increase the amount, knead, and dilute and solidify. radiation
As sex nuclides, Co-60, Cs-137, Tc-9
9, Ni-59, Ni-63, Sr-90, I-12
9, Nb-94, C-14, H-3, transuranium element
At least one is included.

The present invention can be carried out even if the sampling line and the measuring / measuring unit are installed before or after the dehydrator. As the solidifying material of the present embodiment, water-hardening cement and cement glass are suitable, but thermoplastic plastic and asphalt can also be used.

Next, a volume reduction system for non-combustible miscellaneous solid wastes such as metal pipes, concrete, heat insulating materials, etc. generated from a nuclear facility will be described with reference to FIGS.

FIG. 3 shows that the solid waste is compressed and reduced in volume, filled in a solidification container, and solidified by pouring the solidified material. The miscellaneous solids stored in the storage container 11 are taken out by turning over the container, sorted for each single item, and placed on the transfer conveyor 12. A dose rate meter 13 collimated with a shielding material is attached in the middle of the conveyor, and measures the dose rate at that position for each single product. The waste that has passed the management set value is put into the secondary container 14, and the waste that exceeds the set value is transferred to the storage container 15 by another route. At this time, if the secondary container is vibrated, the filling rate is improved. After the secondary container is filled with the waste, the container is pressed by the press 16 together with the container. The press machine is preferably one capable of reducing the diameter so as to open a gap between the press body 17 and the solidification container 19, but it is also possible to make the secondary container smaller in size than the solidification container.
After the weight and radioactivity are measured by the weighing / measuring unit 18, the press body 17 is labeled with a number so that it can be compared with the measurement result, and is temporarily stored. Radioactivity measurement
This is carried out by installing a NaI scintillator or a semiconductor detector configured in advance using a standard radiation source, connecting a multi-channel analyzer, and obtaining a counting rate for each nuclide. In this example, Co-60 and Cs-137, which are γ nuclides, are measured,
Other α nuclides and β nuclides are evaluated by multiplying the previously described scaling factor by a safety factor. After that, two or three press bodies are filled in the solidification container 19, and at this time, a combination of press bodies that does not exceed the standard for burying the solidified body with respect to the concentration of the radionuclide is selected. After filling the press body,
The solidified material is poured into the solidified container from the mixer 20 to form a solidified body.

In this embodiment, a measuring / measuring unit is installed at the front stage of the press machine to measure the weight and the nuclide concentration of each of the miscellaneous solids. In this way, the type and weight of miscellaneous solids supplied to the press may be controlled. In this case, the dose rate meter installed on the conveyor can be omitted.

Next, high-temperature melting will be described with reference to FIG. The method shown in FIG. 4 is to melt miscellaneous solid waste at a high temperature (1600 ° C. or higher at which iron or concrete melts), fill it into a solidification container, inject the solidified material, and solidify it.
The method of taking out the iron material waste is the same as in the above embodiment.
The miscellaneous solids stored in the storage container 21 are supplied to the melting furnace 24 using the transfer conveyor 22. A measuring / measuring unit 23 is installed at the exit of the conveyor, and the weight and radioactivity of each waste product are measured, and then the waste is put into a melting furnace.
The weight and nuclide concentration of the input waste are stored and recorded in the computer 25 connected to the measuring / measuring unit, and the cumulative amount of one batch is monitored. The radioactivity measurement is carried out by installing a NaI scintillator or a semiconductor detector configured using a standard radiation source in advance and connecting a multi-channel analyzer to obtain a counting rate for each nuclide. In this example, Co-60 and Cs-137, which are γ nuclides, are measured, and other α nuclides and β nuclides are evaluated by multiplying the scaling factor described above by a safety coefficient. Through this process, it is possible to evaluate the nuclide concentration of the melt 26 in the furnace using the measured values of the weight of the input material and the nuclide concentration stored in the computer. The type and amount of miscellaneous solids to be introduced are controlled so that the nuclide concentration of the melt does not exceed the standard value for burying the solidified material. Therefore, it is desirable that the waste having a high radioactivity level in the measurement stage in the measuring / measuring unit be separated and stored. The melt thus obtained is poured into a refractory container 27 smaller in size than the final solidification container, and is allowed to cool before filling into a solidification container 29. The solidification material is injected into the gap between the refractory container and the solidification container from the mixer 28, and the formation of the solidified body is completed.

In this embodiment, the melt is dropped into water,
It may be cured as granular slag. The obtained slag can be directly mixed with cement and solidified. Since the nuclide concentration in the melt is below the reference value, the solidified body does not exceed the reference value. Further, a method is also possible in which after measuring the weight and radioactivity of each single waste product, filling the refractory container and supplying the entire container to a melting furnace.

The solidified waste of the miscellaneous solid waste thus prepared is finally checked for nuclide concentration by an inspection facility.
According to the present embodiment, it is possible to eliminate the generation of solidified material that exceeds the burial management reference value for miscellaneous solid waste whose radioactivity level is distributed over a relatively wide range.

Another embodiment of the present invention will be described with reference to FIG. The present embodiment relates to a system for incinerating combustible miscellaneous solid waste. The measuring / measuring unit 31 of the present invention is installed at the front stage of the incinerator 30 to measure the weight and the nuclide concentration of the single solid product taken out of the storage container 32. 1/500 reduction in volume ratio due to incineration in consideration of safety factor
Then, even if the waste has a concentration of 1/500 of the burial management reference concentration, the incinerated ash reaches the reference concentration. Therefore, the operation reference value is set three digits below the buried management reference concentration.
As a result of the measurement, those exceeding the operation standard value are separated and transferred to the solidification facility 33 to be solidified or stored again. Only the waste that passes the operation standard value is incinerated, and the incinerated ash is mixed with the solidified material in a solidification facility and solidified in the solidification container.

According to this embodiment, it is possible to eliminate the generation of solidified material exceeding the burial control standard value, to prevent the incinerator and the filter equipment from being contaminated, and to improve the maintainability of the equipment. It is also effective in reducing the exposure of workers.

[0027]

According to the present invention, the event that the concentration of radionuclides in the solidified waste finally solidified or the surface dose rate of the solidified solid exceeds the burial control standard value can be prevented beforehand, and the waste can be stored again. And the need for additional measures such as dilution and solidification.

Further, it is useful for preventing contamination of the processing equipment, and is also effective for maintenance of the equipment and for reducing the exposure of workers.

[Brief description of the drawings]

FIG. 1 is a diagram showing a processing flow of a radioactive waste liquid according to an embodiment of the present invention.

FIG. 2 is a diagram showing a flow of a radioactive waste resin solidification processing system according to an embodiment of the present invention.

FIG. 3 is a view showing a flow (compression reduction) of a non-combustible miscellaneous solid volume reduction processing system according to an embodiment of the present invention.

FIG. 4 is a view showing a flow (high-temperature melting) of a non-combustible miscellaneous solid volume reduction processing system according to an embodiment of the present invention.

FIG. 5 is a diagram showing a flow of an incineration system for combustible miscellaneous solids according to an embodiment of the present invention.

[Description of Signs] 7, 19, 29: solidified container, 8: inspection device, 9: sampling line, 10, 18, 23, 31: measuring / measuring unit, 13: dosimeter, 16: press machine, 17: press Body, 24: melting furnace, 25: computer, 26: melt, 27
... refractory container, 30 ... incinerator, 33 ... solidification equipment.

──────────────────────────────────────────────────の Continued on the front page (51) Int.Cl. ⁷ Identification code FI G21F 9/30 571 G21F 9/30 571E 9/32 9/32 Z (72) Inventor Tsukasa Baba 7-chome, Omikamachi, Hitachi City, Ibaraki Prefecture No. 2 Hitachi Energy Co., Ltd. (72) Inventor Koichi Chino 7-2-1 Omika-cho, Hitachi City, Ibaraki Prefecture Hitachi, Ltd. Energy Research Laboratory (72) Inventor: Jin Kikuchi Omika-cho, Hitachi City, Ibaraki Prefecture 7-2-1 Hitachi Energy Co., Ltd. (72) Inventor Tatsuo Izumida 3-1-1 Kochicho, Hitachi-shi, Ibaraki Pref. Hitachi, Ltd. Hitachi Plant (56) References JP-A 62-100700 (JP, A) JP-A-59-97099 (JP, A) JP-A-3-81700 (JP, A) (58) Fields investigated (Int. Cl. ⁷ , DB name ) G21F 9/36 G21F 9/30

Claims (5)

(57) [Claims] 1. A measuring the radioactivity information of radioactive waste, the processing method of the radioactive waste is determined based on the radiation information, processing the radioactive waste on the basis of the determined processing method A radioactive waste treatment method, wherein the determined treatment method is such that the radioactivity information is lower than a reference value.
The radioactive waste and the radioactivity
Mix with radioactive waste whose information is lower than the reference value
A method for treating radioactive waste, which is a treatment. 2. The radioactive waste treatment method according to claim 1, wherein the volume of the mixed radioactive waste is reduced, and the volume-reduced radioactive waste is filled in a solidification container and solidified. 3. The radioactive waste treatment method according to claim 1, wherein said radioactivity information is a concentration of radionuclides contained in said radioactive waste. Wherein said volume reduction process, the compression volume reduction, melting, method for treating a radioactive waste according to claim 2 or claim 3 is either incinerated and dehydration. 5. The radioactive waste according to claim 1, wherein said radioactive waste contains at least one of a waste liquid, a concentrated waste liquid, a noncombustible miscellaneous solid waste, an ion-exchange resin and a waste resin. Waste treatment method.