JPS6361639B2 - - Google Patents

Description

[Detailed description of the invention]

[Technical Field of the Invention] The present invention relates to a method for solidifying radioactive waste, and more specifically, it includes pulverizing a vitrified body obtained by heating and melting a mixed powder of a radioactive waste calcined body and a glass-forming substance, The present invention relates to a method for obtaining a solidified radioactive waste having excellent corrosion resistance in high-temperature and high-pressure water by adding ZrO ₂ -Al ₂ O ₃ -rare earth oxide to this pulverized material, melting it, cooling and solidifying it. [Technical background of the invention and its problems] With the spread of nuclear power generation, the amount of highly radioactive waste fluid generated from spent nuclear fuel reprocessing plants is increasing year by year. Since storing them poses safety issues, there is a strong desire to establish a conversion technology to solid storage bodies that can be stored more safely. Generally, when disposing of radioactive waste, the waste is reduced in volume and solidified, and the resulting solidified material is thermally, chemically, and mechanically stable, and even during long-term storage, radioactive materials are released into the outside world. It is requested that the spread of the virus be minimized as much as possible. From this point of view, conventional solidification methods are based on vitrification technology (for example,
-3240 and 50-4840) are the mainstream. That is, the high radioactivity level waste liquid, which is usually generated as a nitric acid solution, is first calcined to remove the evaporated components, and the high radioactivity level waste is separated in the form of nitrate or oxide powder. This is a method in which the powder is melted with a glass system such as phosphoric acid or borosilicate glass, and then solidified by solidifying into an ingot of a certain shape. According to the above method, by considering the waste content, glass composition, etc., it is possible to obtain a vitrified material with relatively high mechanical strength, but it has the following problems. In other words, 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 the temperature inside the vitrified body rises to 500 to 700°C at the center of the solidified body. It may extend to. Therefore, for example, if vitrified material is stored in a geological formation and comes into contact with water, the vitrified material will be eroded by high-temperature water that has become hot due to the heat released from itself. Therefore, safe and stable storage of radioactive waste over a long period of time becomes difficult. In particular, it is known that when a vitrified material is eroded by high-temperature water, a layered substance is formed on the surface of the solidified material (for example, S., Pickering; Journal of American
Ceramic Society magazine, Vol. 63, No. 9-10, p. 558,
(1980), the film formed in this way is extremely fragile and easily peels off, which has the disadvantage of further dispersing radioactive materials to the outside world. [Object of the Invention] An object of the present invention is to solve the above-mentioned drawbacks of the conventional vitrification treatment method for radioactive waste, and to provide a method for obtaining a solidified material with excellent leaching resistance against high-temperature water. [Summary of the Invention] According to the research conducted by the present inventors, the solidified material obtained by the conventional vitrification treatment method for radioactive waste contains
It has been found that by further adding Zr-Al-rare earth oxide, the erosion resistance and leaching resistance against high-temperature water can be dramatically improved. The radioactive waste solidification treatment method of the present invention is based on such knowledge, and more specifically, 50 to 80% by weight of the vitrified material containing radioactive waste has a chemical composition of (i) ZrO. ₂ 40 to 80% by weight, (ii) 30 to 10% by weight of Al ₂ O ₃ , and (iii) 30 to 10% by weight of rare earth oxide.
It is characterized by melting at a temperature of ~1600°C and solidifying upon cooling. [Specific Description of the Invention] The present invention will be described in more detail below. In the following description, "parts" and "%" are based on weight unless otherwise specified. Radioactive waste to be treated by the present invention includes, for example, U,
In addition to the remaining radioactive waste that has been collected, various kinds of waste liquids containing radioactive materials, such as concentrated recycled waste liquid from mixed-bed desalination equipment, concentrated waste liquid from floor drains or equipment drains generated from buildings, etc. Various types of solid waste such as used ion exchange resin and filter sludge generated from the reactor water purification system, fuel pool system, condensate system, and drain system, and precipitated sludge generated from coagulation and sedimentation treatment of waste liquid, etc. Contains high-level and medium-low level radioactive waste. By calcining these radioactive wastes,
A calcined body of radioactive waste is obtained. Such a radioactive waste calcined body is dispersed or co-melted in a melt of glass-forming components such as borosilicate glass, phosphate glass, etc. in an amount of 70 parts to 30 parts thereof, according to conventional methods, It is considered to be vitrified. Currently, vitrified solidified radioactive waste is being produced in this way, and the amount accumulated is gradually increasing. The present invention aims to convert the vitrified material once produced and stored into a more stable solidified material by adding 40 to 80% _ZrO2 to the vitrified material.
Al ₂ O ₃ 30-10%, rare earth oxides (e.g.
It is characterized by adding an additive consisting of 30 to 10% of Sm ₂ O ₃ , Y ₂ O ₃ , Ce ₂ O ₃ , Nd ₂ O ₃ ), melting it, and then cooling and solidifying it. Regarding the mixing ratio of the vitrified material and the additive, if the amount of the vitrified material is too small, the waste treatment efficiency will decrease;
If the amount of the additive is too small, the leaching resistance of the resulting solidified product will be reduced, so the range of the additive is preferably 50 to 20% with respect to 50 to 80% of the vitrified product. In addition, although a borosilicate-based vitrified material is preferably used as a raw material, it is not necessarily necessary to grind it into a powder, and it may remain in the form of granules or lumps as long as it does not impede melting and mixing. The mixture of the vitrified material and additives is melted by a conventional method such as high-frequency induction heating, but if the melting temperature is less than 1300°C, the viscosity is high and it is difficult to obtain a uniform mixture. If the temperature exceeds 1600℃, the evaporation loss of Na etc. will be significant, so 1300℃
A range of ~160°C is preferred. Then, by cooling this melt, a solidified body according to the present invention is obtained. It is not necessarily clear why the addition of additional additives to the vitrified body obtained by the conventional method dramatically improves the leaching resistance of the solidified body as shown in the Examples. However, when the solidified material obtained by the present invention was examined by X-ray diffraction, it was confirmed that a crystalline substance was partially formed in the solidified material, and the presence of this crystalline material was found to be a factor in the leaching resistance. It is thought that this contributes to the improvement of [Examples and Comparative Examples of the Invention] Example 1 Calcined powder of simulated radioactive waste having the composition shown in Table 1 below (simulating the oxide obtained by calcining waste liquid from a reprocessing plant) was Prepared.

[Table] After uniformly mixing the above-mentioned simulated radioactive waste calcined powder and the glass-forming material powder with the composition shown in Table 2 below at a weight ratio of 3:7, the mixture was placed in a platinum crucible and heated to 1300℃. Melted.

[Table] This was then cooled to room temperature to obtain a vitrified simulated radioactive waste. The vitrified material was crushed into powder. 50% by weight of the vitrified powder of simulated radioactive waste obtained in this way,
30% by weight of ZrO ₂ powder, 10% by weight of Al ₂ O ₃ powder and rare earth oxides (Ce ₂ O ₃ : 2.0% by weight, Nd ₂ O ₃ : 2.0
After uniformly mixing 10% by weight of powder (Sm ₂ O ₃ : 6.0% by weight), the mixture was placed in a platinum crucible and heated to 1500° C. in an electric furnace to melt it. Next, the melt was allowed to cool and solidify in an electric furnace together with the platinum crucible. It was confirmed by X-ray diffraction that some crystalline substances were precipitated in the obtained solidified body. Next, a cube-shaped test piece with a side of 10 mm was cut out from the solidified body, inserted into an autoclave, and kept in high-temperature water under equilibrium vapor pressure of 300 ° C. for 24 hours.
The leaching properties of the components contained in the solidified body were investigated (hereinafter referred to as leaching test). The weight change of the test piece before and after the leaching test was measured, and the leaching rate of the solidified material (the amount of components contained in the solidified material leached out per unit surface area of the solidified material in 24 hours) was calculated, and the leaching rate was 1.67× It was 10 ^-3 g/cm ² day. This value is based on the leaching rate in the comparative example described below.
The leaching resistance is approximately 10 times higher than that of 1.60×10 ⁻² to 1.90×10 ⁻² g/cm ² ·day. Examples 2 to 6 A vitrified powder of simulated radioactive waste obtained in the same manner as in Example 1 was prepared, and a mixed powder having the powder composition shown in Table 3 below was placed in a platinum crucible and heated in an electric furnace. After melting, the mixture was cooled to obtain a solidified product. It was confirmed from the results of X-ray diffraction that crystalline substances were precipitated in all of the obtained solidified bodies. Cubic test pieces with a side of 10 mm were cut out from these solidified bodies, and a leaching test was conducted under the same conditions as in Example 1. The leaching rate is shown in the right column of Table 3 below.

[Table] In Table 3 above, the composition of the rare earth oxides is as shown in Table 4 below.

〔Effect of the invention〕

As is clear from the above-mentioned Examples and Comparative Examples,
The solidified radioactive waste obtained by the method of the present invention has about 3 to 10 times improved leaching resistance against high-temperature water compared to the solidified material obtained by the conventional method.
Suitable for long-term storage of radioactive waste.

Claims (1)

[Claims] 1. Vitrified material containing radioactive waste 50 to 80
50% by weight of additives with a chemical composition of (i) 40-80% by weight of ZrO ₂ , (ii) 30-10% by weight of Al ₂ O ₃ , (iii) 30-10% by weight of rare earth oxides. A method for solidifying radioactive waste, comprising mixing ~20% by weight, then melting the mixture at a temperature of 1300 to 1600°C, and solidifying by cooling.