KR20190075599A - Volume Reduction and Vitrification Treatment Method for Spent Uranium Catalyst Waste without Generation of Secondary Wastes - Google Patents

Abstract

The present invention relates to a uranium waste immobilization processing method capable of reducing volume of uranium waste catalyst up to 1/4, satisfying uranium concentration limit of intermediate/low level radioactive waste without generating secondary waste by adding a vitrification agent into the uranium waste catalyst to vitrify the uranium waste catalyst. The method comprises: a step (1) of mixing uranium waste catalyst with the vitrification agent to manufacture silicate or borosilicate having low melting temperature; and a step (2) of charging the mixture into a crucible to melt and vitrify the same. According to the present invention, by mixing uranium waste catalyst with the vitrification agent and performing a melting process in a crucible, it is possible to reduce volume up to 1/4 such that a domestic disposal site to be constructed in the future is used effectively and cost for disposing of uranium waste catalyst can be reduced.

Description

Technical Field [0001] The present invention relates to a method for reducing and securing uranium waste catalysts in which secondary wastes are not generated,

The present invention relates to a method for treating uranium waste catalysts free from secondary wastes. More particularly, the present invention relates to a method for treating waste uranium catalysts, which comprises the steps of immobilization of uranium spent catalyst to be disposed of, The present invention relates to a method for treating spent uranium catalysts in which no waste is generated.

Radioactive waste should be isolated from the external environment for a long period of time at a specific repository until the risks are neglected due to the high risks to the natural environment, taking into account the radiation half-life period. Radioactive waste or radionuclides must be immobilized by a variety of materials in order to prevent the outgassing of hazardous materials from the radioactive waste to the natural environment for a long period of time. At this time, depending on the physicochemical properties of the radioactive waste, cement, plastic, asphalt, The material is solidified using materials such as glass and ceramics.

In Korea, acrylonitrile, synthetic fiber materials since the 1970's acrylic composite oxide of USbO ₅ of uranium (U) and antimony (Sb) to:: (SiO ₂ substrate) (Acrylonitrile CH ₂ = CHCN) in the porous silicon oxide to produce Or USb ₃ O ₁₀ supported uranium catalysts have been developed. Uranium of this uranium catalyst uses only depleted uranium with almost all of uranium-235 removed and almost uranium-238 only.

  In Korea, these uranium waste catalysts have to be treated as radioactive waste based on the Atomic Energy Act. In the case of a domestic private company, acrylic synthetic fibers using uranium catalyst were produced until 2004, and about 7,100 drums (1 drum = 200 liters) of uranium spent catalyst were generated and stored.

According to Article 205-18 of the Ministry of Internal Affairs and Communications, the radioactivity standard value of radioactive waste is less than 3,700 Bq / g, which is 14.6 parts by weight of uranium in case of natural uranium, Uranium content corresponds to 25.2 parts by weight. In order for these radioactive wastes to be released to the environment at the level of non-radioactive waste, uranium in the solid waste must have a very low uranium concentration of less than about 0.005 part by weight.

At present, the radioactivity of uranium waste catalyst using the depleted uranium generated in Korea is in conformity with the acceptance criteria of radioactive waste repository to be operated in Gyeongju in the future, but the disposal cost is about 8.5 million won per 200L drum, It is estimated that the final disposal waste will reach nearly 10,000 drums, considering the volume increase due to additives in the process of manufacturing solid waste for the disposal of uranium waste catalyst, Except for this, direct disposal costs can reach 100 billion won.

Since uranium used in the uranium waste catalyst has little economic value as the depleted uranium, if the uranium is selectively selectively disposed of in the radioactive waste repository, the volume of waste to be disposed can be theoretically up to about 95% The disposal cost is about 5 billion won. However, it is technically very difficult to chemically separate uranium from Fe, Si, Sb, and other elements contained in the uranium spent catalyst, and it is advantageous in terms of reducing the amount of uranium, but the acidic solution or alkaline The generation of secondary wastes from the solution and uranium that is not completely separated in solution will act as a new source of pollutants. Also, even if uranium is completely separated from uranium spent catalyst, it has a problem of exceeding the uranium content of the regulation for the middle and low level radioactive waste delivery.

Therefore, there is no generation of secondary wastes in view of the effective use of the domestic repository to be constructed and the reduction of the cost of disposal of uranium waste catalysts, and the technology for screening uranium waste catalysts that does not exceed the uranium content of the regulations on the delivery of low- and mid- It is a technology that urgently needs to be developed at the national level.

A conventional method for depressurizing uranium waste catalyst is a method in which sodium compounds such as sodium carbonate (NaCO ₃ ) and sodium hydrogen carbonate (NaHCO ₃ ) are mixed with a uranium spent catalyst as disclosed in Patent Document 10-0587157, after reacting in made of sodium silicate (Na ₂ SiO _3), to this mixture and the temperature of 10 ~ 200 ℃ water to produce a liquid sodium silicate, to separate the solid material of a liquid sodium silicate and is not dissolved in the solid-liquid separator Method. In Patent Registration No. 10-1207339, an uranium waste catalyst is reacted with an alkali carbonate at a high temperature to form an alkali salt, which is then dissolved in water. A carbonate is added to the prepared aqueous solution to extract uranium and recover it as metal uranium. In Patent Document 10-926462, a uranium catalyst is heated at a high temperature to separate volatile substances such as antimony oxide (Sb ₂ O ₃ ) and molybdenum oxide (MoO ₃ ), and vitrification agents are added to the remaining waste in which volatile substances are separated and removed And melting at high temperature to produce vitrified uranium solidified body. U.S. Patent No. 9,181,602 B2, in which uranium spent catalyst is dissolved in an alkali solution, and the undissolved material is dissolved in an acidic solution. Then, pH is controlled by using carbonate and hydrogen peroxide to separate silicon (Si) and uranium (U).

The above-described conventional uranium processing technology has low economic efficiency due to a combined process apparatus such as a high-temperature melt reaction process, a high-temperature high-pressure liquid phase reaction process, a centrifugal separation, and a high-temperature volatilization process. In addition, existing technologies for separating silicon (Si) from aqueous solution and solidifying residual uranium require separate treatment of radioactive liquid waste due to the residual uranium in the liquid waste. In particular, the uranium waste catalyst wastes must satisfy the domestic uranium concentration limit (less than 3,700 Bq / g, less than 14.6 parts by weight of natural uranium, and less than 25.2 parts by weight of depleted uranium) of low- and mid-level radioactive waste. However, due to the high uranium content of the uranium spent catalyst, it is impossible to remove all of the silicon oxide because of exceeding the uranium concentration limit of the middle and low level radioactive waste when all the silicon oxide is removed. However, most of the existing technologies do not include all the unnecessary silicon and silicon oxide We are trying to separate.

The Applicant has found that the addition of a vitrification agent to the uranium waste catalyst liberates the uranium spent catalyst to liberate secondary wastes, satisfies the uranium concentration limit of the low- and mid-level radioactive wastes, To provide a possible method.

Korean Registered Patent No. 10-0587157 (Registered on May 29, 2006) Korean Registered Patent No. 10-0926462 (Registered on Nov. 05, 2008) Korean Registered Patent No. 10-1207339 (Registered Nov. 27, 2012) US registered patent US 9181605 B2 (Nov. 10, 2015)

It is an object of the present invention to provide a volumetric and immobilization treatment method of spent catalyst containing uranium to effectively absorb the volume of uranium spent catalyst without generating secondary wastes and satisfy the domestic standard of low and mid level radioactive waste repository And a method for treating uranium spent catalyst.

In order to accomplish the above object, the present invention provides a method for producing a mixture comprising: (1) mixing a uranium waste catalyst of a solid powder with a vitrification agent to prepare a mixture; And (2) charging the mixture into a crucible, melting and vitrifying the mixture, wherein the vitrification agent is an immobilization treatment method of uranium waste which is an alkali metal carbonate, an alkali oxide, a borate, a boron oxide, As an aspect of the present invention.

The uranium spent catalyst in the step (1) may be in the form of a compound represented by the following general formula (1) supported on a silicon dioxide (SiO ₂ ) support.

[Chemical Formula 1]

U _w Sb _x M _y O _z

(M = at least one selected from among Fe, Al, Mo, V and Bi, and w, x, y and z are molar ratios constituting the oxide).

The uranium (U) concentration of the uranium spent catalyst in the step (1) may be 0.005 to 25 parts by weight based on 100 parts by weight of the uranium spent catalyst.

The alkali metal carbonate may be at least one or more of Na ₂ CO ₃ , K ₂ CO ₃ , CaCO ₃ and MgCO ₃ .

The alkali oxide may be at least one or more of Na ₂ O, K ₂ O, CaO and MgO.

The borate salt may be at least one or more of NaBO ₂ , Na ₂ B ₄ O ₇ , KBO ₂ and CaBO ₃ .

In the step (1), a mixture prepared by mixing 5 to 20 parts by weight of the vitrification with respect to 100 parts by weight of the uranium waste catalyst may be used.

In the step (2), the heating temperature may be 1000 to 1300 ° C, and the melting may be performed for 1 to 4 hours.

It is another aspect of the present invention to provide a uranium waste solidified body having a compressive strength of 18 to 23.5 MPa produced by the method for immobilizing uranium waste.

According to the present invention as described above, uranium-containing radioactive waste as well as uranium-containing waste catalysts can be simultaneously disposed for volume and immobilization through a single process, and can be disposed of as low- and mid-level radioactive waste.

In addition, the present invention is economically feasible by treating radioactive waste through a single process, does not generate secondary wastes, and is suitable for domestic low-level radioactive waste repository acceptance standards (3,700 Bq / g or less, 14.6 wt% 25.2 wt% or less of depleted uranium), and at the same time, the volume of waste is reduced to 1/3 to 1/4.

Fig. 1 (a) is a photograph showing a mixture of uranium waste and a vitrification agent before treatment, Fig. 1 (b) is a photograph showing a mixture of uranium waste and a vitrification agent, It is an image of a solidified body.

Hereinafter, the present invention will be described in detail.

According to one aspect of the present invention, there is provided a method for reducing and immobilizing uranium waste catalyst, comprising the steps of: (1) mixing a uranium waste catalyst and a vitrification agent to form a mixture; And (2) heating the mixture.

The uranium spent catalyst in the step (1) is not limited to the uranium spent catalyst for producing acornitrine, and the uranium spent catalyst may be a form in which a compound represented by the following formula 1 is supported on a silicon dioxide (SiO ₂ ) support .

[Chemical Formula 1]

U _w Sb _x M _y O _z

(M = at least one selected from among Fe, Al, Mo, V and Bi, and w, x, y and z are molar ratios constituting the oxide).

The uranium spent catalyst of the step (1) includes silicon dioxide (SiO ₂ ), antimony oxide, uranium (U) and iron oxide, and the uranium (U) The concentration may be 0.005 to 25 parts by weight relative to 100 parts by weight of the uranium spent catalyst.

Preferably, in the uranium spent catalyst, the silicon dioxide is contained in an amount of 44 to 55 parts by weight based on 100 parts by weight of the uranium spent catalyst, 25 to 30 parts by weight of the antimony oxide, 5 to 10 parts by weight of the uranium, And the iron oxide may be 1 to 5 parts by weight, and the antimony oxide may be one or a combination of two or more selected from the group consisting of Sb ₂ O ₃ , Sb ₂ O ₄ and Sb ₂ O ₅ , , Fe ₂ O ₃ and Fe ₃ O ₄ .

The vitrification agent in step (1) may be one or more selected from the group consisting of alkali metal carbonates, alkali oxides, borates, boron oxide (B ₂ O ₃ ) and aluminum oxide (Al ₂ O ₃ ) May be a mixed oxide of two or more kinds, and may be a compound capable of forming a silicate or borosilicate glass having a low melting point. More specifically, the alkali metal carbonate is at least one or more of Na ₂ CO ₃ , K ₂ CO ₃ , CaCO ₃ and MgCO ₃ , The alkali oxide is at least one or more of Na ₂ O, K ₂ O, CaO and MgO, and the borate may be at least one of NaBO ₂ , Na ₂ B ₄ O ₇ , KBO ₂ and CaBO ₃ .

In the step (1), one or two or more of the vitrification agents may be added to the uranium spent catalyst in order to form vitrification, and the content thereof is preferably 20 parts by weight based on 100 parts by weight of the uranium spent catalyst. More preferably 10 to 15 parts by weight. If the amount of the vitrification agent is less than 5 parts by weight, melting of the uranium waste catalyst may not occur and it may be difficult to produce a vitrified solidified body, and if it exceeds 20 parts by weight, the production cost may increase.

In the step (2), the mixture prepared in the step (1) may be charged into a graphite crucible for heating the mixture, and then melted in a box-shaped electric furnace. In order to form a uranium waste catalyst solidified body, Is preferably carried out at 1000 to 1300 ° C for 1 hour to 4 hours, more preferably at 1100 to 1250 ° C for 2 hours to 3 hours.

The uranium waste solid body produced by the above method has a compressive strength of 18 to 23.5 MPa, and the initial volume of uranium spent catalyst is reduced to 1/3 to 1/4.

In addition, the uranium concentration is characterized by satisfying the disposal limit of domestic low-level radioactive waste (3,700Bq / g or less, 14.6wt% based on natural uranium, 25.2wt% or less of depleted uranium).

Hereinafter, the present invention will be described in more detail with reference to Examples. It is to be understood by those skilled in the art that these examples are for illustrative purposes only and that the scope of the present invention is not construed as being limited by these examples.

Example 1.

Referring to Figure 1 (a), silicon dioxide (SiO ₂₎ of the solid powder has to as a form supported on the support a composition of Table 1 Catalyst uranium (MAC-3, Solutia Inc.) 90 parts by weight of sodium carbonate (Na ₂ CO ₃ ) was charged into a graphite crucible. Then, the graphite crucible was charged into a box-type electric furnace and the experiment was conducted. The heating rate of the electric furnace was 5 ° C./minute and maintained at 1200 ° C. for 4 hours for melting and solidification. After the experiment, the solidified material was cooled in an electric furnace.

Furtherance The Content (wt%) 실리콘 이산화물 SiO ₂ 45-55 안 미 산화물 _{Sb 2 O 3, Sb 2 O} 4, Sb 2 O 5 25-30 Uranium U 5-10 Iron oxide _{FeO, Fe 2 O 3, Fe} 3 O 4 1-5

Example 2.

The uranium waste catalyst solidified body was prepared in the same manner as in Example 1, except that the experiment was carried out at 1150 ° C for solidification in Example 1.

Example 3

The uranium waste catalyst solidified body was prepared in the same manner as in Example 1, except that the experiment was carried out at 1250 ° C. for solidification in Example 1.

Example 4

A uranium spent catalyst solidified body was prepared in the same manner as in Example 1, except that 10 parts by weight of potassium carbonate (K ₂ CO ₃ ) was used as a vitrification agent in Example 1.

Example 5

A uranium spent catalyst solidified body was prepared in the same manner as in Example 1 except that 10 parts by weight of boron oxide (B ₂ O ₃ ) was used as a vitrification agent in Example 1.

Example 6

A uranium spent catalyst solidified body was prepared in the same manner as in Example 1 except that 10 parts by weight of a mixture of sodium carbonate (Na ₂ CO ₃ ) and boron oxide (B ₂ O ₃ ) was used as a vitrification agent in Example 1.

Measurement example

The physical properties of the uranium spent catalyst solidified bodies prepared in Examples 1 to 6 were measured by the following methods.

(1) Compression test

The compressive strength of the uranium waste catalyst solidified body according to the embodiment was measured by using a tensile tester (MTS-810) according to the ASTM C39 method after processing the uranium solidified body to a diameter of 10 mm and a height of 20 mm.

(2) Radiation leaching experiment

The activity of the uranium waste catalyst solidified body according to the embodiment was measured by measuring the radioactivity of water after immersion for 10 days in water at 30 to 35 ° C.

The measured values of physical properties of Examples 1 to 6 in which the measurement examples were carried out are summarized in Table 2 below.

Compressive strength (MPa) Radioactivity detection (Bq / ml) Volume Reduction Rate (%) Tolerance 3.44 MPa or more 0.08 Bq / ml or less Example 1 22.2 0.069 67.7 Example 2 18.7 0.073 62.9 Example 3 23.5 0.058 69.7 Example 4 20.05 0.075 65.5 Example 5 19.7 0.081 60 Example 6 21.6 0.077 64.2

Having described specific portions of the present invention in detail, those skilled in the art will appreciate that these specific embodiments are merely preferred embodiments and that the scope of the present invention is not limited thereby. something to do. Accordingly, the actual scope of the present invention will be defined by the appended claims and their equivalents.

Claims (9)

(1) mixing the uranium waste catalyst of the solid powder with a vitrifying agent to prepare a mixture; And
(2) charging the mixture into a crucible, melting and vitrifying the mixture,
Wherein the vitrification agent is an alkali metal carbonate, an alkali oxide, a borate, a boron oxide, an aluminum oxide or an oxide mixed with them.
The method according to claim 1,
Wherein the uranium spent catalyst in the step (1) is a form in which a compound represented by the following formula (1) is supported on a silicon dioxide (SiO ₂ ) support.
[Chemical Formula 1]
U _w Sb _x M _y O _z
(M = at least one selected from among Fe, Al, Mo, V and Bi, and w, x, y and z are molar ratios constituting the oxide).
The method according to claim 1,
Wherein the uranium (U) concentration of the uranium spent catalyst in the step (1) is 0.005 to 25 parts by weight relative to 100 parts by weight of the uranium spent catalyst.
The method according to claim 1,
Wherein the alkali metal carbonate is at least one of Na ₂ CO ₃ , K ₂ CO ₃ , CaCO ₃ and MgCO ₃ .
The method according to claim 1,
Wherein the alkali oxide is at least one or more of Na ₂ O, K ₂ O, CaO and MgO.
The method according to claim 1,
Wherein the borate salt is at least one of NaBO ₂ , Na ₂ B ₄ O ₇ , KBO ₂ and CaBO ₃ .
The method according to claim 1,
The method for immobilizing uranium waste using the mixture prepared by mixing 5 to 20 parts by weight of the vitrification with respect to 90 parts by weight of the uranium spent catalyst.
The method according to claim 1,
Wherein the step (2) is performed at a temperature of 1000 to 1300 캜 for 1 to 4 hours.
A uranium waste solid body having a compressive strength of 18 to 23.5 MPa produced by the method for immobilizing uranium waste according to claim 1.

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