JPS61132898A - Method of solidying and treating radioactive waste - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

DETAILED DESCRIPTION OF THE INVENTION [Technical Field of the Invention] The present invention relates to a method for solidifying a radioactive concentrated waste liquid of medium to low level containing sodium nitrate generated in a nuclear fuel reprocessing facility.

[Technical background of the invention 1ll Generally, spent nuclear fuel is generated as a result of the operation of a nuclear power plant, and this spent nuclear fuel contains U or Pu.
It contains valuable materials that can be used as fuel for fast breeder reactors such as, and these valuable materials are recovered at spent nuclear fuel reprocessing facilities.

Recovery of U or Pu from spent nuclear fuel is usually performed using P.
This is performed using a method called the urex method.

In this method, spent nuclear fuel is dissolved in nitric acid to ionize U or Pu, and this solution and tributyl phosphate (
(hereinafter referred to as TBP)-n dodecane mixed organic solvent is contacted and recovered.

In other words, U or pu ions tend to form complexes with TBP depending on temperature, acid concentration, and valence, so they are made to form complexes using TBP, transferred from a nitric acid solution to an organic solvent, and then mixed with water. By bringing them into contact with each other, U or Pu is back extracted and recovered by adjusting the valence and acid concentration.

The recovery rate can be increased by performing the above process in several stages in series.

This reprocessing process generates waste liquid containing TBP degraded by nitric acid and radiation irradiation, as well as waste liquid from decontaminating building floors. Depending on the level of radioactive waste liquid, they are handled differently as follows.

In other words, the high-level radioactive waste liquid is mainly the waste liquid from the spent nuclear fuel melting process and the first stage extraction process (radioactivity content of 103G+/♂ or more), and as is well known, it is the waste liquid that comes from the spent nuclear fuel melting process and the first stage extraction process (radioactivity content of 103G+/♂ or more), and as is well known, it is a waste liquid that is generated from the spent nuclear fuel melting process and the first stage extraction process. They are mixed and so-called vitrified.

On the other hand, medium to low-level radioactive waste liquid contains low concentrations of radioactivity, nitric acid, and sodium hydroxide. This radioactive waste liquid is reduced in volume by coagulation sedimentation or evaporative concentrator, and is stored in medium-low level radioactive waste liquid storage tanks. It is said that when 1 ton of spent nuclear fuel is processed, 200 to 300 m2 of medium-low level radioactive waste liquid is generated, and safe processing of these medium-low level radioactive liquid waste liquids is desired.

As methods for treating this low-level radioactive waste liquid, currently being researched are the asphalt solidification method and the plastic solidification method.

The asphalt solidification method is a method in which low-level radioactive waste liquid or dry powder of low-level radioactive waste liquid is supplied into asphalt that has been heated to a temperature of 100 to 200°C to form a fluid state, and the radioactivity is fixed in the asphalt. .

On the other hand, the plastic solidification method is a method in which low-level radioactive waste liquid is dried and solidified, often using thermosetting resin. All of these methods have the advantage that the water in the radioactive waste liquid evaporates, resulting in greater volume reduction.

[Problems in the Background Art] However, both the asphalt solidification method and the plastic solidification method were developed for the solidification of low-level radioactive waste fluid generated from nuclear power plants, and this method is not applicable to nuclear fuel. There were the following problems in applying this method to waste liquid generated from reprocessing facilities.

In other words, although the mechanical and chemical properties of the solidified material produced by these asphalt solidification methods or plastic solidification methods satisfy all the conditions required for solidified radioactive waste, Because the solidifying material is an organic material, there is a problem that it deteriorates if kept at high temperatures for a long time.

On the other hand, when looking at the durability of a solidified body from the type of radioactivity contained in the solidified body, the main nuclear ladder of low-level radioactive waste generated from nuclear power plants is ■C0, and its half-life is approximately 5 years. It is.

Therefore, if the solidified material is managed for 50 years, the radioactivity will decrease by 1/2.
In 1000.100 years, it becomes 1/1000000.

In contrast, most of the radioactivity contained in radioactive waste fluid generated from nuclear fuel reprocessing facilities is fission products, including those with extremely long half-lives. for example,"
Since the half-life of Cs is 30 years, the radioactivity can be reduced to 1/10.
It takes 300 years to reduce it to 00, and 600 years to reduce it to 1/1000000. Therefore, over such a long period of time, it cannot be guaranteed that there will be no crustal deformation or heating of the solidified material due to fire, and therefore, strict management over a long period of time is required, which requires a considerable amount of expense.

In addition, a general problem with the asphalt solidification method and the plastic solidification method is that both only involve drying as pretreatment of waste, and there is no process to decompose the waste itself, so it is difficult to reduce the volume further. There is a problem that it becomes inconvenient when required.

[Purpose of the Invention] As a result of intensive research aimed at eliminating the drawbacks of the asphalt solidification method or plastic solidification method, the present inventors have succeeded in reducing the amount of sodium nitrate generated in facilities that handle radioactive materials such as nuclear fuel reprocessing facilities. We have discovered that by drying the concentrated waste liquid containing medium to low level radioactivity and heating and melting it together with a vitrification material, a vitrified material with a high volume reduction rate and excellent mechanical and chemical properties can be obtained.

That is, medium- and low-level radioactive waste fluid generated from nuclear fuel reprocessing facilities contains sodium carbonate in addition to sodium nitrate and trace amounts of sodium nitrite. The melting points of these sodium compounds are 308°C, 308°C,
The temperature is 852°C, and the temperature becomes WJaz above this temperature. Each of these sodium compounds is a material used in the glass industry, and is decomposed and vitrified by reaction with silicon oxide.

This vitrified material has excellent durability and is ideal for long-term containment of radioactivity.

The present invention was made based on this knowledge, and an object of the present invention is to provide a method for solidifying radioactive waste generated from nuclear fuel reprocessing facilities and the like.

[Summary of the Invention] As mentioned earlier, medium-low level radioactive waste liquid (radioactivity concentration of 103 CI/IN3 or less) generated from nuclear fuel reprocessing facilities contains sodium carbonate in addition to sodium nitrate and sodium nitrite. However, these sodium compounds each have the following (1) in the presence of silicon dioxide:
It reacts as shown in formula (2) to become soda-silicate glass and is vitrified.

2Na N0342Na NO2+022Na NO2
-+Na 20+N2 +('1/2)02Na 20
+Si02-+Si02-Na20...(1) Na2CO3-+NazO+CO2Na20
+SiOz-+SiO2 ・NazO...(
2) The reactions (1) and (2) proceed easily at about 1200°C.

Further, when boric acid, aluminum hydroxide, and calcium carbonate are used in combination with silicon dioxide as a vitrification agent, the vitrification reaction can proceed smoothly.

In other words, if only silicon dioxide is used as a vitrifying agent, the viscosity of the resulting glass will increase, making it difficult to form a homogeneous glass, and the melting temperature will gradually increase, but using boric acid will increase the viscosity of the glass. The temperature decreases, and the melting temperature also begins to decrease.

In addition, the glass formed according to the present invention has a high soda content in order to improve volume reduction properties, and therefore has low water resistance, but boric acid and aluminum hydroxide have the effect of improving the water resistance of the glass. do.

As mentioned above, most of the chemical components contained in medium- and low-level radioactive waste fluid generated from nuclear fuel reprocessing facilities can be used as raw materials for glass, so radioactive waste fluid is dried to form radioactive powder, and these powders are Add silicon dioxide powder
When melted in Lagoon B, the sodium compounds in the powder are easily decomposed into sodium oxide, which reacts with silicon dioxide and becomes vitrified.

What should be noted here is that in the present invention, sodium nitrate itself is decomposed to release nitrogen and w1 elements, resulting in a significant reduction in the amount of the product. Furthermore, the radioactivity in the radioactive concentrated waste liquid is confined within the glass body, resulting in an extremely safe solidified product.

By the way, if this soda-silicate glass contains more than 20% by weight of sodium oxide, its water resistance becomes extremely poor.
The melt is known as water glass. It's a shrimp.
It is also known that when the sodium oxide content exceeds 50118%, the glass state cannot be maintained.

Therefore, the present invention creates a sodium borosilicate glass by adding boron as a third element and an aluminum compound and a calcium compound as other components, and improves various properties such as water resistance and meltability of the glass body formed. It is.

In other words, the radioactive waste solidification treatment method of the present invention converts solid or liquid medium- to low-level radioactive waste containing sodium nitrate generated at a nuclear fuel reprocessing facility into silicon dioxide, boron compounds, aluminum compounds, and alkaline earth. It is characterized in that it is heated and melted together with similar metal compounds to produce glass solid waste containing at least 1511 iffi% of Na2O incorporating radioactive compounds.

The radioactive waste used in the present invention may be obtained by concentrating and drying a radioactive waste liquid to form a powder, or may be used as a concentrated waste liquid. However, in the case of concentrated waste liquid, the heat required for melting is used to evaporate the water, which increases the time required to make glass.
Further, problems such as scattering of the melt occur. Therefore, it is preferable to use the radioactive waste in the form of a dry powder.

The amount of silicon dioxide used in the present invention is suitably 25 to 65% by weight, particularly 35 to 45% by weight of the total compounded product. If it is less than 25% by weight, the water resistance of the solidified product will be poor, and if it exceeds 65g1% by weight, the volume reduction properties will be poor and the melting temperature will also be high.

Examples of the boron compound used in the present invention include boric acid, borax, and dehydrated borax, among which boric acid is particularly preferred. Boron compounds other than boric acid, especially borax and dehydrated borax, which introduce 1/2 mole of sodium to boron in atomic ratio into the vitrified body are not preferred.

The blending amount of the boron compound is 1 to 20% in terms of boric acid in order to increase the meltability of the vitrified material and improve the water resistance.
Weight % is appropriate.

Examples of the aluminum compound used in the present invention include aluminum hydroxide, alumina (AJ!zOs), and kaolin (Abu203.2SiOz.2H20), among which aluminum hydroxide is particularly preferred. The amount of aluminum compound added improves the water resistance of the vitrified material.

Furthermore, in order to prevent devitrification, it is appropriate to add 1 to 5% by weight of the total blending amount in terms of aluminum hydroxide. Addition of 5% by weight or more is not preferable because it significantly reduces the meltability of the glass.

The alkaline earth metal compounds used in the present invention include:
Examples include calcium compounds such as calcium carbonate, dolomite (Ca CO3-vo CO3), etc.
Among these, calcium carbonate is particularly preferred.

The appropriate amount of the calcium compound is 1 to 15% by weight of the total amount in terms of calcium carbonate.

In addition to the main component, sodium nitrate, the dry powder of medium- and low-level radioactive waste generated from nuclear fuel reprocessing facilities contains crud consisting of iron oxides and hydroxides produced by corrosion of metal parts of piping. However, according to the present invention, it has been experimentally shown that this cladding component also becomes a stable oxide (a metal oxide whose main component is Fe203) and is incorporated into the glass body. ii! It has been certified.

The amount of radioactive dry powder added in the present invention is appropriately set so as to obtain desired water resistance, volume reduction properties, and meltability.

That is, by adjusting the amount of radioactive waste liquid dry powder added, it is possible to design a product with water resistance comparable to that of a vitrified body of high-level waste liquid.

The sodium borosilicate glass obtained by the present invention is quite different in composition from industrially produced sodium borosilicate glass such as Pyrex glass (manufactured by Corning Glassworks), and has the following characteristics. .

In other words, sodium oxide, which is the main glass auxiliary raw material, is obtained by decomposing sodium compounds in radioactive waste liquid, and its content is at least 15 fflffi%, which is higher than other glasses, and it contains radioactivity. Therefore, it is necessary to improve the intensities as much as possible by containing a boron component and an alkaline earth metal represented by calcium oxide, and the total amount of the boron oxide component, alkaline earth metal oxide component, and soda component is m of silicon dioxide. It has characteristics such as being very large compared to the glass body, and adding an aluminum compound to maintain meltability and increase the water resistance of the glass body.

In general, compared to glass used for high-level radioactivity concentrated waste liquid treatment, it has a lower melting temperature (
It has the characteristics of low viscosity and good meltability.

[Embodiments of the invention] Examples of the present invention will be described below.

Example 1 Dry powder of radioactive waste fluid generated at a nuclear fuel reprocessing facility (12% sodium nitrate, 28% sodium carbonate) was mixed with silicon dioxide, boric acid, aluminum hydroxide, and calcium carbonate in the proportions shown in Table 1. The mixture was mixed and heated and melted at about 1200°C. Note that the numbers in the table below indicate weight %.

The composition of the glass thus obtained was as shown in Table 2.

(Left below) Example 2 Sodium nitrate 28% by weight, sodium carbonate 28% by weight
A simulated dry powder of nitrate-containing acid waste liquid generated at each fuel reprocessing facility was prepared by mixing the following: Next, this simulated dry powder was mixed with 50i1 ffi% of silicon dioxide, 13% by weight of boric acid, and 111% of aluminum hydroxide. % and calcium carbonate 7 songs ω%
After mixing, heat at 1200℃ to create a glass solid,
We investigated its physical properties.

The results are shown in Table 3.

(The following is a blank space) [Effects of the Invention] As explained above, according to the present invention, radioactive glass solid waste has water resistance, heat resistance, little deterioration due to aging, and excellent chemical and mechanical strength. can manufacture things.

Representative Patent Attorney Suyama Sa Continued from page 1 of education

Claims (9)

[Claims] (1) Solid or liquid medium-low level radioactive waste containing sodium nitrate is heated and melted together with silicon dioxide, boron compounds, aluminum compounds and alkaline earth metal compounds, and at least 15 weight of Na_2O containing radioactive compounds is added. % or more of glass solid waste. (2) The solid radioactive waste containing sodium nitrate is radioactive dry powder obtained by evaporating radioactive concentrated waste liquid containing sodium nitrate generated at a nuclear fuel reprocessing facility. radioactive waste solidification treatment method. (3) The method for solidifying radioactive waste according to claim 1 or 2, wherein the boron compound is boric acid. (4) The method for solidifying radioactive waste according to any one of claims 1 to 3, wherein the aluminum compound is aluminum hydroxide. (5) The method for solidifying radioactive waste according to any one of claims 1 to 4, wherein the alkaline earth metal compound is calcium carbonate. (6) The blended amount of silicon dioxide is 25 to 65 of the total blended amount.
% by weight. The method for solidifying radioactive waste according to any one of claims 1 to 5. (7) The method for solidifying radioactive waste according to any one of claims 1 to 6, wherein the amount of the boron compound is 1 to 20% by weight of the total amount in terms of boric acid. . (8) Solidification treatment of radioactive waste according to any one of claims 1 to 7, wherein the amount of the aluminum compound is 1 to 5% by weight of the total amount in terms of aluminum hydroxide. Method. (9) The radioactive waste according to any one of claims 1 to 8, wherein the amount of the alkaline earth metal compound is 1 to 15% by weight of the total amount in terms of calcium carbonate. Solidification treatment method.