KR101865353B1 - Method for vitrifying radioactive rare earth waste - Google Patents

Abstract

The present invention relates to a method for vitrifying radioactive rare earth waste.
As described above, according to the present invention, it is possible to stably vitrify the radioactive rare earth waste by injecting up to 20 wt% of the radioactive rare earth waste by using the glass medium containing alumina, Can be produced. In addition, the possibility of the liquid-liquid phase transition occurring when treating only the borosilicate glass system is also basically prevented by using the base glass system containing alumina. In view of the estimated liquid temperature and transition temperature, Since the solid phase is not only stable in the process, or during disposal, the glass phase is not likely to be transferred to the glass-ceramic containing the crystal phase, but is less than 0.1 ° C, which generates heat at the final generated glass solid surface, There is also an effect that it is possible to increase the acceptance of residents in the above repository.

Description

TECHNICAL FIELD [0001] The present invention relates to a radioactive rare earth waste,

The present invention relates to a method for vitrifying radioactive rare earth waste, and more particularly to a method for vitrifying radioactive rare earth waste using a glass medium containing alumina.

Spent fuel refers to the time at which a nuclear fuel in a nuclear power plant loses its capacity as a nuclear fuel after it has reached its end of its electricity production life. The nuclear fuel to be charged to the nuclear power plant is uranium oxide, and the U-235 concentration reaches about 4.5 wt%, and the remaining about 95.5% of the uranium consists of U-238 which does not fission. After about four years of electricity production at the reactor, it will emit from the reactor, burning for four years, with about 1.2 wt% plutonium (Pu) and heavier than uranium, radiation and a radiation half- (Np, Am, Cm, etc.) of about 0.21 wt%, and radiation does not emit much, but has a half-life that can penetrate nature and contaminate the soil over many years I-129 and Tc-99 elements are about 0.16 wt%, and the half-life for releasing radiation is not so long. However, Cs and Sr, which emit a large amount of radiation and generate heat, contain about 0.53 wt% About 4.95wt% of Fission Product, and about 92.95wt% of stable elements including residual uranium not participating in fission. Therefore, the spent fuel has 7.05wt% of elements generated in the process of nuclear fission, making it very difficult for people to handle or access it directly. Therefore, it is necessary to safely isolate the spent fuel in our living environment, It is our reality to have homework to make you feel not to feel.

In order to expand the use of nuclear energy, it is a prerequisite to manage the spent fuel generated from nuclear power plants in a safe and environmentally friendly manner and to solve the problem of depleted uranium resources. For this purpose, major nuclear-power advanced countries are concentrating their efforts on pyroprocessing, which is a technology to dry spent nuclear fuel. Korea started to develop pyro-processing technology in 1997. This technology is electrolytic treatment of spent nuclear fuel by using electricity at a temperature of 500 ° C or higher in a hydrochloric acid salt state almost similar to melting salt. The technology first converts spent fuel into metal. This metal contains plutonium, including uranium, and a small group of nuclear materials that have a long half-life and emit much radiation. If uranium is used again in hot molten salt medium, only uranium can be selectively recovered. Therefore, it is possible to separate the elements that have a long half life, which is the problem of spent fuel, and radiate a large amount of radiation, and it is easy to store them in a small scale even after separation, so that these elements are no longer affected by our environment It is very suitable for treatment. Then, electricity is again used to collect a small amount of nuclear material, including residual uranium and plutonium. Therefore, this technology is fundamentally blocked by the selective separation of plutonium as seen in past reprocessing techniques, and even with an artificial method, the electrochemical properties of each material, including uranium, It is a technology that can not isolate any nuclear fuel material by itself. In other words, it is a technology that attracts attention in terms of nuclear nonproliferation because it is a technology that can not separate only plutonium.

Waste of pyro-processing can be broadly divided into metal waste and salt waste. In the case of metal waste, it is a waste generated from cladding hull, unheated fission products and other process equipment. In the case of salt waste, LiCl salt waste containing a large amount of Cs, Sr, etc. from the electrolytic reduction process is discharged from the electrolytic refining process, Lt; RTI ID = 0.0 > LiCl-KCl < / RTI >

The rare earth chloride nuclides present in the electrolytic refining eutectic salt wastes are separated and removed from the eutectic salt in the form of oxychloride or oxide in the form of stable nuclear species through reaction with oxygen. The rare earths that are formed by the oxychloride (REOCl) are Eu, Gd, Sm, La, Nd and Pr, and the rare earths which precipitate in oxide form are Ce, Pr (REO2), Y (RE2O3). Pr is formed of both oxychloride and oxide. The precipitates are mixed with a structurally small cubic oxide and a large tetragonal oxychloride. Rare earth oxides isolated during the treatment of salt wastes are stable materials and their heat generation and radioactivity are not so high, so that researches are underway to solidify or solidify ceramics. Research on the production of solid waste of radioactive waste composed only of powdery rare earth oxide separated during the treatment of salt waste has been carried out by Korea Atomic Energy Research Institute (KAERI).

KAERI has been engaged in the production of rare earth oxide solid by the R7T7 glass medium, which is a borosilicate glass system used for the production of glassy solid of high-level waste in France, and solid state and monazite ceramics based on borosilicate glass system after synthesis of rare earth oxides with rare earth monazite (RE- The rare earth oxide solid was prepared by the ZIT (Zinc Titanate) solidification medium composed of CaHPO4-ZnO-TiO2-SiO2-B2O3 as the solidification medium and the leaching and physicochemical properties were compared and evaluated.

However, these solidification processes can produce excellent solidification when they are performed precisely in small-scale laboratories. However, considering the actual process, there is a problem in practicality due to the complexity of the process. In addition, since the level of waste is still high-level waste with a certain amount due to the decontamination coefficient of U and TRU of about 2,000 in the rare-earth waste, the complicated treatment process should be applied, and various devices necessary for the process should be added and installed . Considering the acceptance of residents in Korea for the treatment and disposal of high-level waste, it is a reality that even if it is processed in the process, it can not guarantee disposal of solid waste to the environment.

In order to overcome this problem, we have increased the decontamination factor of U and TRU (about 2900) in rare earth waste generated from conventional pyrolysis process by more than 10 times, and have succeeded in converting pyrolysis of all high level waste generated in spent nuclear fuel recycling process A verification study of the PyroGreen process is being conducted. It is suggested that vitrification of rare earth waste generated in this process is the most optimal. However, in the case of the base glass component, there is a possibility of causing a liquid-liquid phase separation in which two glass phases are present in one glass due to the ratio of B and Na, and therefore it is necessary to improve the base glass composition And the final glass solid is at the midpoint level and appropriate measures should be presented to identify the relevant properties.

On the other hand, related prior arts include Korean Patent Laid-Open No. 10-2010-0133089 (a method for producing a ceramic solidified body containing a radioactive rare earth oxide, and a ceramic solid body having improved density, thermal stability, and anti-erosion property).

The present inventors prepared a glass solid using a glass medium containing Al ₂ O ₃ , B ₂ O ₃ , Li ₂ O, Na ₂ O, and SiO _2, and found that the solid phase homogeneity, leaching resistance, The amount of heat generated, and the like.

(2) introducing a glass medium containing alumina into the melting furnace to vitrify the radioactive rare earth waste, and (3) introducing the molten glass into the melting furnace. Melting the radioactive rare earth waste and the glass medium; And (4) solidifying the molten glass. The present invention also provides a method for vitrifying a radioactive rare-earth waste.

Another object of the present invention is to provide a glass solidified body comprising a glass medium containing radioactive rare earth waste and Al ₂ O ₃ , B ₂ O ₃ , Li ₂ O, Na ₂ O and SiO ₂ .

In order to accomplish the above object, the present invention provides a method for producing radioactive rare earth waste, comprising the steps of: (1) introducing radioactive rare earth waste into a melting furnace; (2) introducing a glass medium containing alumina into the melting furnace to vitrify the radioactive rare earth waste; 3) melting the radioactive rare earth waste and the glass medium; And (4) solidifying the molten glass. The present invention also provides a method for vitrifying a radioactive rare earth waste.

In the step (1), the radioactive rare-earth waste is selected from Sc, Eu, Gd, Tb, Dy, Ho, Er, La, Ce, Nd, Pm, Sm, Tm, Y oxides and Ac, Th, Pa, , Am, and Cm.

In the step (2), the glass medium containing alumina includes Al ₂ O ₃ , B ₂ O ₃ , Li ₂ O, Na ₂ O, and SiO ₂ .

In the step (3), the radioactive rare earth waste is contained in an amount of 1 to 20% by weight, and the remainder is a glass medium.

The 3 free medium in step is Al ₂ O ₃ 1 to 5 wt%, B ₂ O ₃ 10 to 13 wt%, Li ₂ O 1 to 5 wt%, Na ₂ O 20 to 30 wt%, SiO ₂ 30 By weight to 40% by weight, and the balance being radioactive rare earth waste.

In the step (3), the temperature is elevated to 1000 to 1400 캜 and melted.

The present invention also provides a glass body comprising a glass medium comprising Al ₂ O ₃ , B ₂ O ₃ , Li ₂ O, Na ₂ O, SiO ₂ and a radioactive rare earth waste.

The glass medium is Al ₂ O ₃ 1 to 5% by weight, B ₂ O ₃ 10 to 13 wt%, Li ₂ O 1 to 5% by weight, Na ₂ O 20 to 30% by weight, SiO ₂ 30 to 40% by weight And the remainder is radioactive rare earth waste.

The radioactive rare earth waste is a mixture of Sc, Eu, Gd, Tb, Dy, Ho, Er, La, Ce, Nd, Pm, Sm, Tm, Y oxides and Ac, Th, Pa, U, Np, .

As described above, according to the present invention, it is possible to stably vitrify the radioactive rare earth waste by injecting up to 20 wt% of the radioactive rare earth waste by using the glass medium containing alumina, Can be produced. In addition, the possibility of the liquid-liquid phase transition occurring when treating only the borosilicate glass system is also basically prevented by using the base glass system containing alumina. In view of the estimated liquid temperature and transition temperature, Not only is the solid state stable during the process or disposal of the glass phase to the glass ceramics containing the crystalline phase, but also because the surface temperature rise of the final glass surface is insignificant to less than 0.1 ° C and is not within the high-level waste category, It also has the effect of increasing the acceptance of residents in the repository.

Fig. 1 shows the result of the evaluation of the amount of nuclear species, the amount of heat and the radioactivity of the radioactive rare-earth waste.
2 shows the composition of the glass medium according to the invention.
3 is a SEM photograph of the glass according to the present invention.

Hereinafter, the present invention will be described in detail.

In vitrification of rare earth waste such as Y, La, Ce, Pr, Nd, Sm, Eu, and Gd generated in pyroGreen PyroRedSox process, the vitrification process can waste a certain amount of waste, And should be designed to ensure glass quality that meets the disposal criteria. The main process parameters of the vitrification process using the glass composition with a certain amount of waste are the viscosity and the electric conductivity, and the homogeneity, the leaching resistance, the non-radioactivity and the heat generation should be evaluated as the qualities that the glass solid has to have in the disposal environment.

The present invention relates to a process for the production of radioactive rare earth wastes, comprising the steps of (1) introducing radioactive rare earth waste into a furnace, (2) introducing a glass medium containing alumina into the melting furnace to vitrify the radioactive rare earth waste, and (3) Melting the glass medium; And (4) solidifying the molten glass. The present invention also provides a method for vitrifying a radioactive rare earth waste.

The 3 free medium in step is Al ₂ O ₃ 1 to 5 wt%, B ₂ O ₃ 10 to 13 wt%, Li ₂ O 1 to 5 wt%, Na ₂ O 20 to 30 wt%, SiO ₂ 30 By weight of Al ₂ O ₃ , 11.43% by weight of B ₂ O ₃ , 3.43% by weight of Li ₂ O, 25% by weight of Na ₂ O, including%, SiO ₂ 37.00% by weight, and the rest is intended to include radioactive rare earth waste. The melting is preferably carried out by raising the temperature to 1000 to 1400 占 폚, more preferably by raising the temperature to 1200 占 폚.

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.

EXAMPLES Example 1 Evaluation of Nuclear Emission, Heat and Radioactivity of Radioactive Rare Earth Wastes

The pyroGreen process assumes 4.5 wt% U-235 enrichment of spent nuclear fuel (SNF), 45,000 MWd / MTU of combustion, 50 year cooling, U and TRU decontamination coefficient of 20,000, 2, the amount of nuclear species generated, the amount of heat and the amount of radioactivity of the rare earth waste generated by the treatment of spent nuclear fuel 10 MTHM are as shown in Figure 1. As shown in Figure 1, only 99.6 wt% of rare earth waste Nide is included.

Example 2. Determination of composition of glass medium

Much of the calorific value and radioactivity in rare earth waste and Actinide mixtures occur in Am, Pu. The rare earths generated by the bubbling of oxygen in PyroRedSox in PyroGreen process are mostly oxides and are generated in the form of fine powder with an average particle size of about 5 ㎛. The rare earth oxides are stabilized in cubic or hexagonal form, and the melting point of all oxides is higher than 3,500 ° C in the case of Nd ₂ O ₃ , which is the lowest value among the eight oxides, at 2,233 ° C. appear.

The vitrification process and the glass composition for rare earth waste have been considered to be vitrified through cold crucible induction melter (CCIM). When vitrifying rare earth oxides, the minimum solubility should be 20wt%. The viscosity should be 100 poise or less and the electric conductivity should be 0.4 S / cm or more to enable CCIM operation when melting. The glass solid forms should be homogeneous and satisfy the PCT (Product Consistency Test) leaching rate. The free solid is a radionuclide with a heat generation rate of less than 2 kW / m ³ or a radioactivity concentration of more than 20 years, g. < / RTI >

In order to obtain the optimum glass composition satisfying processability and glass quality, glass compositions using eight representative rare earth oxides were evaluated and candidate glass PG14 satisfying various evaluation items was selected (FIG. 2). The composition of the base glass oxide used in the development of the glass composition was Al ₂ O ₃ , B ₂ O ₃ , Li ₂ O, Na ₂ O, and SiO ₂ .

Example 3. Characterization of a glass solidified body according to the present invention

When the rare earth oxides were vitrified in an alumino borosilicate base glass system, the solubility was estimated to be 20 wt%. For reference, the solubility increased with increasing the process temperature and was estimated to be 25 wt% at 1,300 ° C and 30 wt% at 1,400 ° C. The candidate glass showed a viscosity of 7.2 poise and an electrical conductivity of 1.13 S / cm at a melting temperature of 1,200 ℃, and the operating parameters in CCIM were evaluated to be very good. The surface of the candidate glassy solid prepared at the process temperature of 1,200 ℃ was homogeneous (left of Fig. 2, vitrified at 1,200 ℃) and excellent chemical durability. The amount of heat generated for the glassy solid is 3.52 W / m ³ and the radioactivity is 2.72E + 5 Bq / g, which does not correspond to the high-level category. The level of disposal at the WIPP (Waste Isolation Pilot Plant) Intermediate Level Waste).

When the candidate glass (PG14) according to Example 2 was heat-treated at 900 ° C, it was observed that a crystal of Neodymium oxide silicate was formed at the boundary between the crucible and the glass as shown in the right photograph of FIG. . When the candidate resin was heat treated at 950 ℃, the liquidus temperature was estimated to be below 950 ℃ because the glass was homogeneous. Since the temperature at which crystals began to be formed in the glass was 950 ° C or lower and was much lower than the vitrification process temperature (1,200 ° C), the possibility of crystal formation during the vitrification process was evaluated to be low.

The transition temperature of glass was also measured to confirm the presence of homogeneous glass phase transition in the repository environment. The measured transition temperature was estimated at about 400 占 폚. When the candidate glass is disposed in a granite environment of about 500 m underground, the heat generated from the glass is low (<0.1 ° C.) and the temperature of the repository is low (about 30 ° C.) Respectively.

In order to evaluate the chemical durability of the candidate glass, 7day PCT (Product Consistency Test) was conducted in a 40 ℃ DI water environment where the air flow was blocked considering the groundwater temperature and the amount of self - Li, Na, and Si elements were leached out, but the leaching rate was less than 17.5% of the standard value (2 g / m ² ), and rare earth elements including Al were not leached at all. Overall, the chemical durability of glass was excellent Respectively.

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) introducing radioactive rare earth waste into a furnace;
(2) Oil containing alumina for vitrification of the radioactive rare earth waste
Introducing the remainder into the melting furnace;
(3) melting the radioactive rare earth waste and the glass medium at a solubility of 20 to 30 wt% and a melt temperature of 1,000 to 1,400;
(4) solidifying the molten glass; Lt; / RTI &gt;
Wherein the radioactive waste is 10 to 20 wt% and the glass medium comprises 1 to 5 wt% of Al2O3, 10 to 13 wt% of B2O3, 1 to 5 wt% of Li2O, 20 to 30 wt% of Na2O, 30 to 40 wt% Gt; a &lt; / RTI &gt; radioactive rare earth waste.
The method according to claim 1,
In the step (1), the radioactive rare-earth waste is selected from Sc, Eu, Gd, Tb, Dy, Ho, Er, La, Ce, Nd, Pm, Sm, Tm, Y oxides and Ac, Th, Pa, , Am, Cm. &Lt; / RTI &gt;
delete delete delete delete The glass medium comprises 1 to 5 wt% of Al2O3, 10 to 13 wt% of B2O3, 1 to 5 wt% of Li2O, 20 to 30 wt% of Na2O, 30 to 40 wt% of SiO2, and 10 to 20 wt% A glass medium and a radioactive rare earth waste,
The radioactive rare earth waste is a mixture of Sc, Eu, Gd, Tb, Dy, Ho, Er, La, Ce, Nd, Pm, Sm, Tm, Y oxides and Ac, Th, Pa, U, Np, Wherein the glass solidified body is a glass solidified body.
delete delete

Images