JPS6113196A - Glass solidifying treating method of radioactive waste sodium compound - 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] <Industrial Application Field> This invention is applicable to the reprocessing of radioactive waste sodium 11 which has been used in fast reactors and is no longer needed, and the utility generated from spent fuel reprocessing/7Ili facilities. This invention relates to a method for vitrifying waste liquid containing a large amount of sodium.

<Conventional technology> Conventionally, when processing radioactive waste sodium etc. that was used in fast reactors and was no longer needed, the sodium was converted into a stable sodium compound and then mixed with cement, asphalt, etc. to solidify it. However, with this method, the volume reduction ratio is small, and 1q
The drawback is that if the solidified material is buried underground or in water, the radioactive sodium contained within the assimilated material is likely to leach out.

In order to keep the volume reduction ratio relatively high, a technology has been developed to add glass-forming materials such as silicon dioxide and frit to the radioactive waste sodium compound, melt it by heating, and then cool and solidify it into a vitrified material. Although it has been proposed, even in this case, the disadvantage that tIl and resistant sodium are easily leached from the obtained vitrified product cannot necessarily be overcome.

On the other hand, Zircaloy, a zirconium alloy with corrosion resistance and high temperature water resistance, is used for various metal parts of the fuel cladding tubes and fuel assemblies of light water reactors. Since Zircaloy is a high-level metal waste, measures to deal with it are an issue.

<Problems to be Solved by the Invention> Therefore, the present invention is capable of vitrifying radioactive waste sodium compounds to obtain a glass assimilate that is resistant to leaching of sodium, and at the same time, it solves a problem that has long been desired. The purpose of this invention is to provide a treatment method that can effectively treat radioactive Zircaloy waste such as spent fuel cladding tubes.

<Means for solving the problem> That is, the present invention oxidizes radioactive zircaloy waste such as spent nuclear fuel cladding tubes to produce zirconium oxide,
A radioactive waste sodium compound is mixed with a glass forming material and the zirconium oxide, and this mixture is heated to melt and then cooled.

This is a vitrification treatment method for radioactive waste smallium compounds, which is characterized by solidification.

<Specific d2 of the invention> The present inventors have discovered that zirconium oxide (Zr
By adding and mixing O) with the glass forming material,
It is possible to obtain a stable vitrified material that is non-deliquescent and has excellent leaching resistance, and furthermore, as the zirconium oxide used in item 1, radioactive zircaloy waste constituting spent fuel cladding tubes, etc. can be oxidized. This invention was completed based on the discovery that radioactive zirconium oxide, which can be oxidized by V, can be used effectively.

The radioactive waste sodium compounds to be treated in this invention include NaOH, 'Na2O.

Na2Co3. It is a solid or aqueous solution whose main components are Na, No, etc. Sources of such radioactive waste sodium compounds include, for example, radioactive waste sodium from fast reactors and sodium-containing waste liquid from light water reactor reprocessing facilities. The former type of waste sodium has high chemical activity as it is, so an inactivation process is required to convert it into stable sodium compounds such as Na and OH, but in the case of the latter type of sodium-containing waste liquid, sodium is recovered in the waste liquid. Since it is a partially stable sodium compound, there is no need for an inactivation step.

If the radioactive waste sodium compound is an aqueous solution, pretreatment is required to evaporate it and concentrate it to a predetermined concentration.

In addition, radioactive waste sodium compounds such as those mentioned above, which are the materials to be treated, generally contain radioactive substances other than sodium compounds (for example, corrosion products of reactor constituent materials, nuclear fission products of fuel materials, etc.). However, these accompanying radioactive substances can also be stably retained in the vitrified body together with the sodium compound by the treatment method of the present invention.

As the glass forming material used in this invention, glass forming materials conventionally used in the technical field of assimilation treatment of radioactive waste, such as silicon dioxide and frit, can be used. By adding zirconium oxide in addition to the material, leaching of radioactive sodium from the obtained vitrified material can be reduced. In particular, in this invention, the zirconium oxide is radioactive zirconium oxide obtained by oxidizing radioactive zirconium metal used as various metal parts of spent fuel cladding tubes and fuel assemblies. Use it. Oxidation treatment of used Zircaloy
For this purpose, zircaloy can be heated to 900° C. or higher in an oxidizing atmosphere, thereby producing zirconium oxide powder.

The radioactive waste sodium compound to be treated is mixed dry or wet with a glass forming material and zirconium oxide from spent Zircaloy, and then the mixture is preferably melted by microwave heating, cooled, and solidified to form a vitrified product. shall be.

The preferred mixing ratio is that zirconium oxide is in the range of 5 to 20% by weight, the radioactive waste sodium compound is in the range of 30% by weight or less (calculated as Na and O), and the remainder is the glass forming material. When the amount of zirconium oxide is less than 5% by weight, the vitrified material obtained becomes deliquescent,
The leaching resistance of sodium deteriorates, and if the amount exceeds 20% by weight, the degree of eutectic wetting increases when the mixture is melted, and undissolved zirconium oxide tends to remain.

Furthermore, if the sodium compound is mixed in an amount higher than 30% by weight, the amount of undissolved zirconium oxide increases, making it difficult to exhibit the effect of adding zirconium oxide, and the resulting glass assimilate becomes deliquescent.

The heating temperature of the mixture only needs to be the temperature at which each component is eutectic, and although it varies depending on the mixing ratio of each component, it is generally 110
It is in the range of 0 to 1500°C. Note that by adding a melting temperature lowering agent such as boron oxide (8,03) to the mixture as necessary, it is possible to melt the mixture at a relatively low temperature.

By cooling and solidifying the melt obtained from sea urchin, mainly Na, SiO3, Na27r
A glass assimilate consisting of Si 2 O, ZrO 2 alone or a mixture thereof can be obtained.

<Example> Example J5 and a comparative example will be further explained below. Note that all percentages are percentages by weight.

Example 1 Radioactive waste sodium compound, silicon dioxide, zirconium oxide obtained by oxidizing spent zircaloy, and boron oxide as a melting temperature lowering agent were mixed in various proportions shown in the table below, and the mixture was heated in a microwave heating melting furnace. After melting using cold F
JJ, solidified to obtain glass assimilate. The results of examining the solidification state of this vitrified material are shown in the table below.

From this table, the amount of zirconium oxide added is 20% or less,
It can be seen that stable vitrified products (A and B) with no deliquescent property and from which 1-lium is difficult to leach out can be obtained with a sodium compound content of 30% or less (calculated as Na20).

Example 2 A mixture having the following composition was melted, cooled, and melted in the same manner as in Example 1.
It was solidified to form a vitrified product.

Waste NaOH77,49 <Na, O! Atl 6o (+ >S
iO104g λ ZI'022(1'(I B203 35, .6g) The obtained vitrified material had no deliquescent properties and was vitrified with relatively high transparency, so it had an appearance similar to ordinary soda glass. .

On the other hand, for comparison, mixtures (1) and (2) with the following compositions were prepared.
) in the same way! , cool, solidify and prepare a solidified material.

Waste NaOH75.5g 10
0 Q (Na20 conversion p)
(58,50) (77,5
g>Si 0275,50 200 0H
2080m, (12001TIG) All of the obtained solidified bodies had deliquescent properties and did not become vitrified.

<Effects> As explained above, according to the treatment method of the present invention, zirconium oxide is added and mixed with a glass forming material to a soft waste sodium compound and melted.

By cooling and assimilating it, it is possible to obtain a vitrified material that does not easily exude sodium, and it is possible to effectively assimilate radioactive waste sodium compounds.

In addition, since zirconium oxide is obtained by oxidizing radioactive zircaloy waste such as spent fuel cladding, it is effective as a treatment measure for radioactive zircaloy waste. This has the advantage of preventing an increase in

Claims (1)

[Claims] 1. Radioactive zircaloy waste such as spent nuclear fuel cladding tubes is oxidized to produce zirconium oxide, a radioactive waste sodium compound is mixed with a glass forming material and the zirconium oxide, and this mixture is heated and melted. A method for vitrifying radioactive waste sodium compounds, which is characterized by cooling and solidifying the compounds. 2. In the mixing step, zirconium oxide is mixed in an amount of 5 to 20% by weight, waste sodium compound is mixed in an amount of 30% by weight or less (in terms of Na_2O), and the remainder is a glass forming material based on the total weight of the mixture. The method described in paragraph 1.