JPH03264897A - Treatment of high level radioactive waste - Google Patents

Abstract

PURPOSE:To treat high level radioactive wastes without generating a huge amount of secondary wastes and to materialize highly volume reduction of them by heating up the high level radioactive wastes to get a calcined body and then by heating up again at very high temperature and sintering powder of the calcined body of the high level radioactive wastes which are obtained after vaporizing and removing a Cs by heating in a non-oxidative atmosphere. CONSTITUTION:A Cs is vaporized and removed by heating up high level radioactive wastes and by vaporizing a water content and a nitric acid to get a calcined body, and then by heating up again in a nonoxidative atmosphere. By this treatment, restrictions related to synthesis of a solidified body regarding to Cs confinement are eliminated. Then, by heating up the calcined body of high level radioactive wastes to make them a sintered body, high volume reduction is materialized. Since the calcined body has an ability of self sintering, the sintered body can be obtained by putting the calcined body in a vessel and heating it up. Moreover, under higher pressure, a much more dense and stably shaped solidified body can be obtained. As it is not for use as molded goods, 300 atoms as so-called high pressure is more than enough for producing such a densely solidified body.

Description

DETAILED DESCRIPTION OF THE INVENTION [Field of Industrial Application] The present invention relates to a method for treating high-level radioactive waste generated in spent fuel reprocessing processes and the like.

More specifically, this is a treatment method in which Cs (cesium) is vaporized and removed by heating a calcined body of highly radioactive waste in a non-oxidizing atmosphere, and then heated at a high temperature to produce a sintered body with a high range capacity. It is related to.

[Prior Art] High-level radioactive waste generated in the reprocessing of spent fuel by the Burex method is stored in the form of nitrate #I solution containing fission products. In the future, this highly radioactive waste will be solidified by mixing it with a medium such as glass.

In addition to glass, many other materials are being studied as media, including synthetic rock (synrock). The content of fission products in the medium is determined by the solubility of the fission products in the medium, chemical durability (leaching rate in water), removal of decay heat, etc., and is limited to about 10% in vitrification. There is.

Another solidification treatment method is the supercalsine method. This involves adding aluminum compounds to highly radioactive waste to form a solidified material with a high fissile sediment content.

[Problem B to be solved by the invention] In conventional general solidification processing, media such as glass unnecessarily increase the volume of highly radioactive waste. In addition, Cs, which is highly radioactive and has a high calorific value, is solidified without being removed, and the content of fission products is reduced in order to retain Cs (reduce the leaching rate of C3) and remove heat. Attempts have been made to remove Cs by extraction methods, ion exchange methods, etc., which cannot increase the amount of Cs. It is difficult to

The volume of the solidified body should be as small as possible to reduce the costs of its storage and disposal.

To achieve this, it is necessary to increase the content of fission products in the solidified body, but for the reasons mentioned above, it currently remains at about 10%.

In addition, the chemical durability of the super calcine method is poor, and attempts have been made to coat it with metal to improve it.
A significant volume reduction of highly radioactive waste has not yet been achieved.

The purpose of the present invention is to eliminate the drawbacks of the prior art as described above,
It is an object of the present invention to provide a method for processing highly radioactive waste and easily realizing high volume reduction and solidification of highly radioactive waste without generating a large amount of new secondary waste.

[Means for Solving the Problems] The present invention heats highly radioactive waste to form a calcined body, and further heats it in a non-oxidizing atmosphere to vaporize and remove Cs, thereby creating a calcined body of highly radioactive waste with low heat generation. This is a highly radioactive waste treatment method that obtains a calcined powder and sinters the calcined powder by heating it at a high temperature. The features of the present invention are that Cs is removed in advance and that the calcined body is made into a sintered body without being melted.

Highly radioactive waste is usually a nitric acid solution obtained as a raffinate in the spent fuel reprocessing process, and contains almost all the fission products in the spent fuel. The nuclides that are problematic from the perspective of disposal when solidifying highly radioactive waste include transuranium nuclides such as Np and Am, and long-lived nuclides such as Tc and 1, but from the perspective of short-term integrity of the solidified waste, (,
s is the nuclide in question. This is because Cs is an alkali metal and in the form of an oxide is easily soluble in water, highly radioactive, and has a high calorific value.

In the present invention, as shown in FIG. 1, this highly radioactive waste is heated to evaporate water and nitric acid to form a calcined body, and further heated in a non-oxidizing atmosphere to vaporize and remove Cs. This eliminates constraints on solidified body synthesis regarding Cs confinement. Calcination is carried out at a temperature at which moisture and nitrogen oxides can be removed and fission products such as Cs are difficult to vaporize, that is, usually about 600°C and a maximum of 700°C.

For calcining, the rotary kiln method and microwave heating method, which are being studied for the vitrification of highly radioactive waste, can be applied. Some of the fission products include R such as Ru and Mo.
There are elements that exhibit volatility in the oxide state of u0a and Mo5s. In order to suppress this volatilization, it is desirable that the production of these oxides should not proceed as much as possible, and from this point of view, Cs vaporization is performed by heating in a non-oxidizing atmosphere. The heating temperature is usually about 900°C and a maximum of 1200°C.

Next, this calcined powder of highly radioactive waste with low heat generation is heated at high temperature to form a sintered body, achieving high volume reduction and solidification. Since the calcined body has self-sintering properties, a sintered body can be obtained by placing it in a container and heating it. To obtain a sintered body with a better shape, either shape it in advance and then sinter it (pressureless sintering), or shape and sinter at the same time (pressure sintering).
is desirable. Sintering after molding can be performed at normal pressure to take advantage of self-sintering properties, but it can also be performed under pressure to prevent vaporization of nuclides with relatively high vapor pressure. There are no particular restrictions on pressure.

Under high pressure, a denser and shape-stable assimilate is obtained. In this method, since the product is not used as a molded product, a dense solidified product can be produced at a high pressure of 300 atmospheres or less.

It is also effective to add a sintering agent such as a boron compound or a sodium compound to promote sintering. By adding these sintering agents, the density of the sintered body can be improved, but the leaching rate with respect to water becomes worse. Therefore, the amount added should be 20% or less, preferably 10% or less.

The sintering reaction is controlled by atmosphere, addition of reducing agent, temperature, and pressure. Although it is possible to carry out the sintering reaction in air, there is a problem of oxidation of furnace materials, container materials, etc. at high temperatures, which limits material selection. Therefore, in the present invention, it is preferable to carry out the process in an atmosphere of air, nitrogen or argon with a reduced oxygen content, or in a vacuum. Reducing agents are useful for maintaining a reducing atmosphere. Reducing agents include gaseous reducing agents such as hydrogen and carbon monoxide to prevent the creation of new secondary waste, reducing agents that gasify during redox reactions such as carbon,
Reducing agents such as alkaline earth metals and rare earth elements, which are constituent elements of the oxide layer that becomes waste, are used. It is also possible to use a substance such as aluminum that does not adversely affect the oxide phase that becomes waste even if it remains as an oxide. During the sintering reaction, platinum group elements (Ru.

Rh, Pd) may be reduced to a metallic state, but
Since these are stable in a metallic state, there is no problem as a sintered body of highly radioactive waste.

The sintering temperature is 1000-2000°C. The atmosphere, reducing agent, temperature, and pressure for the sintering treatment are appropriately combined depending on the reaction conditions.

Examples of the sintering process are shown in FIGS. 2 to 6, and FIG. 2 is an example of sintering under normal pressure. A sintering container 10 is filled with calcined powder 12 and heated. The powder is sintered and becomes a sintered body 14. When the sintered container 10 is used as a storage container as it is, the sintered container 10 and the sintered body 1 are separated by sintering compaction as shown in the figure.
Since a space is created between the sintered body and the sintered body 4, additional calcined powder is added and the sintering process is repeated to fill as much sintered body as possible as densely as possible. FIG. 3 shows an example of pressureless sintering after acid formation. A molded body of calcined powder 10 is placed in a sintering container 12,
Sinter.

In this case, the sintered body 14 is stored in a separate storage container. t
For the c-type, general methods used in ceramic manufacturing can be applied.

FIG. 4 is an example of pressure sintering. Here, calcined powder 12
A sintering container 10 is filled with the powder, and only the powder is compression-molded in the direction of the white arrow by a press, and then heated to obtain a sintered body 14. FIG. 5 shows another example of pressure sintering, in which the entire sintering container lO is pressurized.

The calcined powder 12 is filled into a sintering container lO, and the sintering container lO is
After sealing the O, degas it. Next, the sintered body 14 is obtained by heating while uniaxially compressing (in the vertical direction) using a press. Furthermore, FIG. 6 is an example of HIP (hot isostatic pressing). Fifth
As in the case shown in the figure, the calcined powder 12 is filled into the sintering container lO, sealed and degassed, and then heated while compressing the sintered container 10 in three rings using the HIP method to obtain the sintered material 14. .

[Effect] Nuclear fission products in spent fuel can be broadly classified into ■metallic elements, ■nonmetallic elements, and ■rare earth elements. Examples of the metal elements include alkaline earth metals, transition metals such as Mo, and platinum group elements. By heating the highly radioactive waste to form a calcined body and further heating, most of the alkali metals in the nonmetallic elements (①) and the metallic elements (②) are removed. They are Sb, T
e, Cs, Rb, etc. As a result, in the case of spent fuel with a burnup of 45,000 MWD/MTU and a cooling period of 5 years, the main components of the calcined body are as follows, excluding elements whose content is 100 g/MTLI or less.

・Alkaline earth metals (Sr, Ba) 3, 3kg/MTLI 8. 7%・Transition metal (
Zr, Mo, Tc)...10. 5kg/MTU 27. 9%・Platinum group elements (Ru, Rh, Pd) 5, 4kg/MTLI 14. 3%・Rare earth elements (Y, La, Ce, etc.)...18.5kg/gate TU49.1% total -37
, 7 kg/MTU By heating and sintering this calcined powder, a solidified body with a high degree of volume loss can be obtained. Incidentally, in vitrification, the weight is 10 times that of nuclear fission products, resulting in a solidified material of 150 J per ton of spent fuel, but in the present invention, the solidified material has a volume of 7.31 J.

[Example 1] The amount of fission products in the spent fuel with a burnup of 45,000 MWO/MTU and a cooling period of 5 years was calculated using the 0RIGEN code, and a simulated waste liquid of a corresponding highly radioactive waste liquid was prepared. This simulated waste liquid was heated to 600°C to form a calcined body, and further Cs was vaporized and removed in an argon atmosphere at 900°C.

45 g of the obtained calcined body was placed in a crucible and heat treated at 1450'C for 1 hour in an argon atmosphere. After cooling,
When the contents were taken out, it was found to be a hardened sintered body. Note that the volume of the sintered body was 9.4 ml. The rate of leaching into water was measured according to JIS-R3502.

The leaching rate is 8 x 10-'g/c1 d, which is about the same level as the vitrified material (5 x 10-'g/cm" d), and has sufficient chemical durability as a highly radioactive solidified material. It was confirmed that there is.

[Example 2] The same operation as in Example 1 was carried out at a heating temperature of 1800°C. The contents are baked hard and the leaching rate is 3 x 1.
It was 0-'g/c-.d.

[Effects of the Invention] In the present invention, highly radioactive waste is heated to form a calcined body, and Cs is vaporized and removed by further heating, so that restrictions such as Cs retention and heat generation due to the presence of Cs can be avoided. Since the calcined body is then sintered, a highly reduced volume solidified body can be produced. Therefore, in the present invention, it is possible to achieve a significant volume reduction and solidification of about 1/20 compared to conventional vitrification. Therefore, it becomes possible to significantly reduce costs in storing and disposing of highly radioactive waste.

Furthermore, since highly radioactive waste is solidified by sintering, corrosion of the solidification container can be reduced, and the material problem that was a major problem in manufacturing highly volume-reduced solidified bodies can be solved. Furthermore, it is possible to perform processing using heater heating, etc., instead of using a special heating method (for example, electron beam heating, plasma heating, etc.), and the processing equipment can be configured easily and at low cost.

Furthermore, in the case of melting, a non-uniform solidified product may be formed due to the crystal structure growth process during the cooling process, but since the present invention uses sintering, a macroscopically uniform assimilated product can be obtained.

[Brief explanation of drawings]

FIG. 1 is an explanatory diagram of the treatment process using the method of the present invention, and FIGS. 2, 3, 4, 5, and 6 are explanatory diagrams of the sintering process, respectively. 10... Sintering container, 12... Calcined powder, 14...
・Sintered body. Figure 1

Claims (1)

[Claims] 1. Highly radioactive waste is heated to form a calcined body, and further heated in a non-oxidizing atmosphere to vaporize and remove cesium to produce a calcined powder of highly radioactive waste with low heat generation. and sintering the calcined powder. 2. The processing method according to claim 1, wherein the sintering treatment is performed under pressure by uniaxial compression or HIP method. 3. The processing method according to claim 1 or 2, wherein a sintering agent is added to the calcined powder during the sintering process.