JPS6022320B2 - How to dispose of radioactive waste - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a method for disposing of radioactive waste, and more specifically, the present invention relates to a method for disposing of radioactive waste, and more specifically, after applying a silicon carbide coating to the surface of granular radioactive waste by a chemical vapor deposition method, the waste is multi-coated with metal and then buried. In particular, the present invention relates to a method for manufacturing a radioactive waste storage body that is chemically and mechanically stable and suitable for semi-permanent storage.

With the spread of nuclear power generation, the volume of highly concentrated radioactive waste fluid generated from spent nuclear fuel reprocessing plants is increasing year by year.

Because there are safety issues in storing these radioactive waste liquids in their liquid state in tanks, there is a need for the first time to establish a technology to convert them into solid storage bodies that can be stored more safely. Generally, when disposing of radioactive waste, the waste is solidified in a form that minimizes the diffusion of radioactive materials into the surrounding area, and the resulting solid storage medium is chemically and mechanically stable and has a long-term lifespan. It is necessary that storage does not cause environmental pollution.

From this point of view, the conventional solidification method is vitrification technology (for example, Jiko No. 46-3240).
, 50-484, etc.) are the mainstream, in which highly concentrated radioactive waste is melted together with glass materials such as phosphoric acid or borosilicate glass, and solidified into glass ingots of a certain shape. , a method of solidifying is used,
The solidified material is then sealed in a storage container and stored, for example in underground storage. According to the above method, waste content,
By considering the composition of the glass, it is possible to obtain a vitrified material with relatively high mechanical strength, but it has the following problems.

If the storage container is damaged, the solidified material will come into direct contact with the external atmosphere, such as groundwater in the case of underground storage, so for long-term stable storage,
It is required to minimize the leaching rate of radioactive materials through water.

In this respect, since the basic material of the vitrified material is glass, which is subject to compositional constraints, it is not necessarily easy to achieve both the mechanical strength or integrity of the solidified material and the reduction of the leaching rate using a glass material. [o} Since the thermal conductivity of glass is inherently lower than that of metals, the temperature inside the vitrified body rises due to the heat generated by the radioactive decay of the radioactive substances it contains, and can reach 500 to 700q0 in the center. obtain.

Such temperature increases weaken the structure of the vitrified body, impair its mechanical and chemical stability, and make safe long-term storage of radioactive waste difficult. As a radioactive waste treatment method that compensates for the drawbacks of the vitrification treatment method described above, a so-called metal matrix method is being considered in which waste radioactive waste or a substance such as a vitrified substance containing it is embedded in metal. .

However, according to research conducted by the present inventors, there are also problems with such a metal matrix method.

The wettability of waste radioactive waste or the vitrified material containing it to metals is not good, and voids may occur in the composite solidified material with these metals, or some kind of thermal shock or mechanical damage may occur. When stress is applied, a phenomenon of peeling may occur at the interface between radioactive waste, etc. and metal. Therefore, the obtained composite solidified material is not necessarily satisfactory in terms of mechanical strength and leaching resistance of radioactive elements. The object of the present invention is to improve the metal matrix method described above and to provide a method for producing a composite solidified storage body that is more stable both mechanically and chemically.

According to the research of the present inventors, in order to achieve the above-mentioned purpose, a silicon carbide film is formed on the surface of granulated materials such as radioactive waste by chemical vapor deposition to improve wettability with metal. It has been found that it is extremely effective to coat the concrete in multiple layers with metal and then bury it.

That is, the method for treating radioactive waste of the present invention includes the step of forming a silicon carbide film with a thickness of 5 to 100 mm on the surface of the radioactive waste or the granular material containing the radioactive waste by chemical vapor deposition. The method is characterized by comprising the steps of: charging the coated granules into an inner container and solidifying them together with the metal; and embedding at least one of the inner containers in the metal in the outer container. The present invention will be explained in more detail below. In the following chronicle, "part" and "
%" is based on weight unless otherwise specified. Radioactive waste to be treated by the present invention includes, for example, radioactive waste remaining after processing spent nuclear fuel and recovering U and Pu, as well as concentrated liquid of recycled waste liquid from hotbed desalination equipment, and Various types of waste liquid containing radioactive substances, such as concentrated waste liquid from floor drains or equipment drains, as well as used ion exchange resins generated from reactor water purification systems, fuel pool systems, condensate systems, and drain systems. It includes high-level and medium-low-level radioactive wastes, such as filter sludge, various solid wastes such as sedimentation sludge produced by coagulation and sedimentation treatment of waste liquids.

By disposing of these radioactive wastes, we can obtain raw materials. Examples of the substance-containing substance include substances obtained by converting the substance into a solidified body using glass, ceramics, metals, etc., and substances in which the solidified body is further subjected to preliminary treatment such as pulverization.

The appropriate content of the katy body in the katy body-containing substance is about 5 to 40%. These crystals or substances containing crystals are molded or crushed to obtain granules. The shape of the granules may be any shape as long as solidification within the inner container is not impaired, but shapes such as a disk, cylinder, sphere, prism or square rod are generally preferably used. There is no particular control on the size of the granules, but if they are too small, it will be difficult to control the thickness of the silicon carbide coating, and if they are too large, the temperature will increase due to radioactive decay in the center. A range of is appropriate. According to the present invention, a silicon carbide film is first formed on the surface of a granular material such as the above-mentioned solid body.

As a chemical vapor deposition method, for example, silicon tetrachloride (SIC) is used.
14) and methane (CH4) in the presence of hydrogen gas at 1100 to 1200 pm to precipitate silicon carbide on the surface of the particles (hydrogen reduction method) or silane (Si ratio) Mixed gas of methane (C ratio) from 600 to
Method of precipitating by thermal decomposition at 120 mm (thermal decomposition method)
Either of these methods may be used. The thickness of the silicon carbide coating is determined to prevent the coating from peeling off or cracking due to stress generated at the interface during cooling to room temperature due to the difference in thermal expansion coefficient between the granules and the coating. , preferably in the range of 5 to 100 Buddhas. The coated granules are then charged into an inner container and solidified together with the metal.

The solidification method is as follows: (a) Compression molding of the coated granules together with metal powder and further scaling if necessary, or charging the coated granules into an internal container and heating the entire container to melt the metal inside; A method of cooling and solidifying, [o' a method of charging the coated granules into an inner container, pouring the molten metal, and then cooling it; conversely, placing an appropriate amount of molten metal in the inner container,
After that, a method is used in which granules are charged and the molten metal is cooled and solidified (in this process, the metal is
It is not necessarily necessary to melt or freeze, but it is necessary to melt or freeze and integrate with the silicon carbide coating of the granules through heating during the subsequent embedding process in the external container).

Metal materials for solidification include Cu, Fe, Pb, Sn, and Zn from the viewpoint of wettability with the silicon carbide film and resistance or meltability within the range of heat-resistant temperature of the inner container material.
, Ni, and alloys containing at least one of these metals as main components are suitable.

For example, when using molten metal in a Cu inner container,
Suitable materials include Pb, Sn, Zn, nails, and alloys containing at least one of these metals as a main component. A suitable content of particulate matter in the solidified material is, for example, about 30 to 50%.

The larger the internal container volume, the more radioactive waste can be solidified, but if it is too large,
This is not preferable because it reduces thermal conductivity and mechanical strength.

For example, in the case of a cylindrical container, the inner diameter is preferably 5 to 5 mm. In addition, the wall thickness of the container is in the range of 0.5 to 5 Yanagi, as the thicker it is, the more durable it is against thinning due to corrosion, and is also advantageous in terms of mechanical strength. is desirable. Next, one or more of the inner containers having the solidified body obtained by solidifying the coated granules therein with the metal are embedded in the metal in the outer container (storage container).

The burying method is essentially the same as the above-mentioned solidification method Kawa-Shiichi, and the "coated granules" are placed in the "inner container".
By replacing "inner container" with "outer container",
The same will be done. However, the number of inner containers is generally smaller than that of granular materials, so even if two or more inner containers are used, they should be spaced appropriately from each other and placed near the center of the outer container to be shielded from the outside world as much as possible. Needless to say, this is desirable. Examples of the material for the inner container or outer container include Fe, Cu, Pb,
Sn, Zn, Ni, Ti, Zr, or an alloy containing at least one of these as a main component is used, and is appropriately determined depending on the solidification method and embedding method.

For example, when a mixture of coated granules and Cu powder is compressed and solidified in an inner container and buried in an outer container using molten Pb, considering the melting point and mechanical strength, Fe , AI, Cu, Nj, Ti, Zr, or an alloy containing at least one of these as a main component is preferably used. After the solidification step and the burying step, a lid made of the same material as the inner container and the outer container is respectively placed on the container, and the periphery is sealed by welding or the like.

The radioactive waste storage body obtained by the method of the present invention as described above has a structure as shown in the drawings, for example, when one inner container is buried in an outer container. That is, granular materials such as radioactive waste coated with silicon carbide 1
is solidified together with the metal 2 in the inner container 3, and the inner container 3 is further embedded in the metal 5 at approximately the center of the outer container 4. As a result of being coated multiple times with metal in this way, the radioactive waste storage body according to the present invention is
It has extremely high mechanical and chemical stability, making it extremely suitable for long-term safe storage. That is, in the case of the storage body according to the present invention, even if the storage container (external container) 4 is damaged, the radioactive waste or solidified granular material 1 will be
Further, since the buried metal 5, the inner container 3, and the solidified metal 2 are shielded from the external atmosphere, for example, seawater, there is extremely little risk of direct contact with the external atmosphere. Considering the extremely low leaching rate of metals into water, long-term safe storage is ensured even in the event of damage to the storage container as described above. Furthermore, even in the unlikely event that water enters through the buried metal 5 and the internal container is damaged, the neighboring radioactive waste granules 1 will be surrounded by the solidifying metal 2 and closely bonded to this through the silicon carbide coating. Therefore, there are almost no voids in the contents of the inner container, and furthermore, peeling at the interface between the granules 1 and the solidifying metal 2 is significantly improved by heat treatment throughout the manufacturing process. Also, due to such a closely integrated structure, the contents of the inner container are
In terms of thermal conductivity, impact resistance, mechanical strength, etc., it is far superior not only to vitrified materials but also to ordinary metal solidified materials. Furthermore, in conventional vitrification treatment methods, the amount of radioactive calcined material that can be stably contained in the glass is limited to 3% by volume;
In the case of the storage body according to the present invention, the content of radioactive calcined bodies can be increased to a maximum of 5% by fracture volume by considering the solidification method, materials, etc. Hereinafter, the present invention will be explained in more detail with reference to Examples and Comparative Examples.

Example A simulated radioactive waste powder (simulating the oxide obtained by evaporating waste liquid from a reprocessing plant) having the composition shown in the table below was prepared.

After uniformly mixing the above-mentioned simulated radioactive waste scale powder and borosilicate glass powder of 100 mesh or less at a volume ratio of 3:7', the mixture was melted at 1200 oo and cooled to obtain a vitrified body. Ta.

A cube with one side and two sides was cut out from this vitrified body.
Place the cube-shaped vitrified body directly into the chemical vapor deposition equipment and apply silane (
A mixed gas of Si ratio) and methane (CH4) is introduced into the device, and a thermal decomposition reaction is carried out for 10 minutes to 1 hour under the conditions of a pressure of 100 tons and a temperature of 60,000 to 8,000○, and silicon carbide is precipitated and coated on the surface of the vitrified material. did. If the thickness of the coating film is less than 5 mounds in a short time, the coating tends to become uneven, and on the other hand, if the thickness exceeds 100 m2 in a long time, the film is likely to crack. Next, as shown in the drawing, vitrified bodies 1 and 1 with good silicon carbide coating (with a film thickness of 5 to 100 μm) were prepared.
Copper powder 2 of 0.00 mesh or less was thoroughly mixed at a volume ratio of 1:1 and placed in a nickel cylindrical container 3 with an inner diameter of 1 mm, wall thickness of 1 inch, and height of 3 mm, and a pressure of 6 tons/me. It was compressed and solidified.

A nickel lid having a diameter of 12 and a thickness of 2 ribs was placed on the nickel container 3, and the surrounding area was sealed with TIC welding. Next, the nickel container is placed in a stainless steel container 4 with an inner diameter of 3 and a height of 5. A heat treatment was performed at 800° C. for 1 hour in an air stream. The storage container was removed from the storage body thus produced to expose the internal solid body.

When this was immersed in 100 kites of pure water at 100°C for 1 hour and a leaching test was performed to detect Cu ions in the solution, the leaching rate was less than the detection limit of 1 x 10-6 days/ground/fem. . Furthermore, the leaching of Mo ions contained in the simulated waste was also less than the detection limit of 1×10-6 evenings/day. Furthermore, a rectangular parallelepiped compression test piece of 5 skins x 5 skins x IQ mosquito was cut out from the center of the nickel container 3 and subjected to a compression fracture test, and the compressive strength was 50k9/part or more, indicating high mechanical strength. Comparative Example 1 The simulated radioactive waste powder used in the example and 10
Volume ratio of borosilicate glass powder of 0 mesh or less: 3:
After mixing uniformly at 7, melt at 120,000,
A vitrified product was obtained by cooling.

Cut out a cube 1 with 2 skins on each side from this vitrified material,
A mixture of copper powder 2 of 100 mesh or less and a volume ratio of 1:1 was placed in a nickel cylindrical container 3 with an inner diameter of 1 m, a wall thickness of 1 m, and a height of 3 m, and a pressure of 6 tons/m. It was compressed and solidified. A nickel lid with a diameter of 12 ribs and a wall thickness of 2 ribs was placed on the nickel container 3, and the periphery was sealed by TIG welding. Next, the nickel container is placed in a stainless steel container 4 with an inner diameter of 3 m and a height of 5 m, and the periphery is covered with copper powder 5 of 100 mesh or less, and then compressed at a pressure of 6 n/m. Sintering treatment was performed at 800oC for 1 hour in a hydrogen stream. The storage container was removed from the storage body thus produced to expose the internal solid body. This is 100
When the Cu ions in the solution were immersed in 100 pieces of pure water for 1 hour, the leaching rate was 1 x 10-6, which was the detection limit.
Evening/The enemy/ship was less than y. Furthermore, the detection limit for Mo ions contained in the simulated waste is 1 x 10-6 days/earth. The ridge was less than y. However, when a rectangular parallelepiped compression test piece of 5 objects x 5 objects x 1 side was cut out from the center of the nickel cylindrical part, a part of the vitrified material peeled off and the compressive strength was less than 5 kg/ground, resulting in mechanical failure. The strength was small.

Comparative example 2 Comparison powder of simulated radioactive waste used in the example and 100
After mixing copper powder with a size below mesh at a volume ratio of 1:1, it is compressed in a mold at a pressure of 6 n/m to form a 10 x 10 x 7.5
It was made into a molded body of ribs.

This molded body was heated in a hydrogen stream for 800°C for 1 hour to undergo a freezing treatment. The leaching rate of Mo ions was determined under the same conditions as in the example and was found to be 4 x 10-3 evenings/day. Comparative Example 3 The competitive powder used in Example 3 and borosilicate glass powder of 600 mesh or less were uniformly mixed at a volume ratio of 3:7, then melted at 1200 oo and cooled to obtain a vitrified product.

A solidified body of 10 x 10 x 7.5 skin was cut out and the leaching rate of Mo ions was determined under the same conditions as in the example.
It was -5 evening/day.

As is clear from the above examples and comparative examples, even if the outermost storage container is damaged and the internal solid body is exposed, the radioactive elements solidified inside the storage body of the present invention can be leached out. It has superior safety and long-term storage properties than simple metal solidified bodies or vitrified bodies.

[Brief explanation of the drawing]

The drawing is a longitudinal cross-sectional view of a storage body obtained by one embodiment of the method of the invention. 1. Particulate matter such as radioactive waste, 2.
...metal for solidification, 3...inner container, 4.
...Outer container, 5...Metal for burial.

Claims (1)

[Claims] 1. A process of forming a silicon carbide coating with a thickness of 5 to 100 μm on the surface of granules of calcined radioactive waste or substances containing the same by chemical vapor deposition, and placing the coated granules in an inner container. A method for disposing of radioactive waste, comprising the steps of charging and solidifying together with metal, and embedding at least one of the inner containers in metal in an outer container.