KR20140058098A - Method of recovering uranium through complexation of polyvinylalcohol-uranium and manufacturing the nuclear fuel pellet by adding polyvinylalcohol-uranium complex - Google Patents

Abstract

The present invention relates to a method of collecting uranium powders and uranium scraps generated during a pellet manufacturing process, uranium ions residing in waste water, and a nuclear fuel pellet manufactured by using the method. According to the present invention, an efficient method for handling the uranium scraps generated during a nuclear fuel manufacturing process and the uranium residing in the waste water is provided. The nuclear fuel pellet, which is manufactured by the uranium handling process and contains polyvinylalcohol-uranium complex serving to act both as a pore forming agent and a U_3O_8 powder, is provided. Therefore, the uranium scraps generated during a nuclear fuel manufacturing process and the uranium residing in the waste water are processed efficiently, the uranium is effectively utilized, and an economic feasibility of the nuclear fuel is improved.

Description

      BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for recovering uranium by a polyvinyl alcohol-uranium complex and a method of producing uranium from the polyvinyl alcohol-

The present invention relates to a method for recovering uranium ions present in uranium scrap and wastewater generated during a process for producing uranium powder and a sintered body, and a nuclear fuel sintered body utilizing the method. More particularly, the present invention relates to uranium scrap and uranium ions (Polyvinylalcohol-uranium) complex, which is capable of simultaneously forming a pore-forming agent and the role of U ₃ O ₈ , is prepared by adding a polyvinylalcohol-uranium complex To a nuclear fuel sintered body.

Nuclear power plant fuel is used in the case of a U-235 to a fissile heavy water 0.7% of the UO ₂ sintered body for light-water reactor is to use U-235 is a UO ₂ sintered body was concentrated to 3-5%. In particular, in the case of a dry method for producing a UO ₂ sintered body for a light water reactor, the UF ₆ in a low concentration state is heated and converted into a gaseous state, which is then converted into UO ₂ powder by reacting with steam or hydrogen. After the UO ₂ powder was converted into granule powder after adding U ₃ O ₈ powder, pore-forming agent and lubricant, the granular powder was compression-molded into a cylindrical shape with a certain diameter and length, The sintered body is ground in a hydrogen atmosphere at 1750 ° C, and the sintered body is ground, washed and dried to complete a sintered body for a light-water reactor. In the process of manufacturing such nuclear fuel, it is possible to produce scrap in the form of solution such as leached uranium, equipment cleaning solution, and the like, which can be generated when the uranium concentration is changed and the equipment is cleaned. And scrap of the form.

At present, ADU processes are mainly used to recycle these uranium scrap. In the ADU process, UO ₂ powder and U ₃ 0 ₈ powder are first dissolved in nitric acid solution to make uranium nitrate (UN) solution, and NH ₃ gas is reacted with the prepared UN solution to form ADU. The formed ADU is filtered, dried and roasted to produce U ₃ O ₈ powder, which is then mixed with UO ₂ powder to produce a sintered body. However, in the method of recovering uranium from conventional uranium scrap, after the ADU powder is prepared, the U ₃ O ₈ powder is roasted and mixed with UO ₂ powder. In the present invention, polyvinyl alcohol (PVA ) chain by coordination bond uranium ions in the UN solution was efficiently recovered uranium and develop new ways high economic efficiency in processing uranium scrap by using a mixture of the U ₃ O ₈ UO ₂ powder directly without roasting the powder in.

Meanwhile, in nuclear power nations, the nuclear fuel cycle technology related to the improvement of nuclear power economy is being developed in various fields. The UO ₂ fuel sintered body, which has been commercialized, has been established to supply nuclear fuel to nuclear power plants. However, in order to utilize uranium efficiently and to improve the economical efficiency of nuclear fuel cycle, Is actively underway.

The nuclear fuel for nuclear power generation generates heat by generating fission in the reactor, and is used in the form of UO ₂ sintered body. UO ₂ fuel has lower thermal conductivity and lower uranium content than metal uranium fuel, but it has a high melting point and is relatively stable to radiation due to nuclear fission.

The UO ₂ sintered body exhibits densification, which is initially reduced in volume when subjected to neutron irradiation, and exhibits a volume swelling in which the volume increases when a certain degree of combustion is reached. If the sintered body has little volume change during burning, it can prevent deterioration of heat transfer due to densification at the initial stage of combustion and reduce deformation of cladding due to volume expansion at the end of combustion. The densification and volumetric expansion have opposite effects on the volume of the sintered body, so that if the two effects are designed to almost cancel out during combustion, there will be little volume change. Since the volume expansion linearly increases with the degree of combustion, the stable sintered body eventually results in changing the microstructure of the sintered body so as to have a densification behavior capable of appropriately compensating for the volume expansion.

It is known that the densification accelerates as the pore contraction decreases as the pore existing in the sintered body shrinks and disappears upon neutron irradiation. The pores of the sintered body before and after combustion show that the pores with diameters of 2 μm or less almost disappear and those with diameters of 10 μm or more are very stable and some of them grow. On the other hand, the volumetric expansion in the furnace is caused by the accumulation of the fission gas product and the fission solid product in the UO ₂ sintered body due to the increase of the sintering body volume as the degree of combustion increases.

The gaseous products in the fission product are hardly dissolved in the sintered body matrix and diffuse to the grain boundaries and then rapidly released to the outside of the sintered body to increase the pressure in the fuel rod to deform the cladding tube. Especially, in the long- . As a method of suppressing the volume expansion of the UO ₂ sintered body in the furnace and suppressing the discharge of the fission gas product outside the sintered body, a space capable of absorbing the fission gas product in the sintered body, that is, a small number of There is a way to make porcelain. As the pore forming agent for making large pores, mainly organic materials such as ammonium oxalate and azodicarbonamide are used.

Currently, UO ₂ sintered body is manufactured by mixing UO ₂ powder with pore forming agent, ADCA (azodicarbonamide), and U ₃ O ₈ powder recovered from uranium scrap. However, the above method does not consider the role of the additive material as a pore-forming agent and the role of the U ₃ O ₈ powder, and may have a problem that the pore-forming agent and the U ₃ O ₈ powder may be present in an uneven distribution have.

Meanwhile, Korean Patent No. 10-0969644 (a method of manufacturing a nuclear fuel sintered body using a high-combustion-grade spent nuclear fuel), Korean Patent No. 10-0569589 (a method of manufacturing a nuclear fuel sintered body ).

It is an object of the present invention to provide a method for efficiently treating uranium scrap generated in a nuclear fuel production process and uranium present in wastewater, And a polyvinyl alcohol-uranium complex which can simultaneously perform the role of a pore-forming agent and the role of a U ₃ O ₈ powder. The present invention also provides a nuclear fuel sintered body produced by adding a polyvinyl alcohol-uranium complex.

Another object of the present invention is to provide a method for producing a nuclear fuel sintered body to which an additive composed of a polyvinylalcohol-uranium complex is added.

(1) preparing a polyvinyl alcohol aqueous solution; (2) dissolving UO ₂ or U ₃ O ₈ powder in nitric acid and diluting with distilled water to prepare uranium nitrate uranium nitrate (UN) solution prepared in step (2), and (3) a solution of uranium nitrate (UN) prepared in step (2) in an aqueous solution of polyvinyl alcohol prepared in step (1) (4) adjusting the pH of the solution prepared in the step (3) to 7 to 11 to form a polyvinylalcohol-uranium complex; And (5) filtering the complex formed by the step (4) and drying the polyvinyl alcohol-uranium complex.

And the pH is adjusted to 7 to 11 by adding NH ₄ OH solution in the step (4).

The present invention also provides a nuclear fuel sintered body produced by adding an additive composed of a polyvinylalcohol-uranium complex to UO ₂ powder in a uranium dioxide fuel sintered body.

Characterized in that 0.05 to 1.0 wt.% Of an additive composed of the polyvinylalcohol-uranium complex is added to the UO ₂ powder.

The present invention also provides a method for producing a uranium oxide fuel sintered body, comprising the steps of: (1) mixing a UO ₂ powder with an additive composed of a polyvinyl alcohol-uranium complex to prepare a mixed powder; ) Preparing a slug by preforming the mixed powder produced in the step (1), and then crushing the slug to prepare a granular powder; (3) compressing and molding the granular powder produced in the step (2) and molding it into a green compact; And (4) sintering the green compact molded by the step (3) under a hydrogen atmosphere.

In step (1), an additive composed of a polyvinylalcohol-uranium complex is added in an amount of 0.05 to 1.0 wt% to the UO ₂ powder.

In the step (3), the granulated powder is compressed and formed into a green compact having a diameter of 10 mm and a length of 11 mm.

And then sintered in a hydrogen atmosphere at 1700 to 1800 ° C for 3 to 4 hours in the step (4).

In accordance with the present invention as described above, it provides a method that the uranium present in the uranium scrap and waste water from nuclear fuel manufacturing processes to handle, have been manufactured through the uranium treatment step, the pore-former Roles and U ₃ O ₈ can be simultaneously treated with a polyvinylalcohol-uranium complex, the uranium scrap generated in the nuclear fuel fabrication process and the uranium present in the wastewater can be efficiently treated , It is possible to contribute to the use of long-term fuel for high combustion by minimizing the volumetric expansion, which increases volume when the volume is initially reduced in the furnace and reaches a certain degree of combustion. As a result, Thereby improving the economical efficiency of the system.

Figure 1 is a process flow diagram of a PVA-U (IV) complex.
FIG. 2 is a process diagram of a UO ₂ fuel sintered body to which a PVA-U (IV) complex is added.
FIG. 3 is a graph showing FT-IR analysis results of PVA and PVA-U (IV) complexes.
Figure 4 is a graph of relative viscosity ratios of PVA and PVA-U (IV) complexes.
Figure 5 is an SEM micrograph of PVA and PVA-U (IV) complexes.
FIG. 6 is a graph showing the X-ray diffraction analysis results of PVA and PVA-U (IV) complexes.
7 is a graph showing the density change of the sintered UO ₂ fuel according to the addition amount of the PVA-U (IV) complex.
FIG. 8 is an optical microscope photograph comparing the pore sizes of the UO ₂ fuel sintered body to which the pure UO ₂ fuel sintered body and the PVA-U (IV) complex are added.
9 is a pure UO ₂ nuclear fuel sintered body and PVA-U: Light microscopic photo of comparing the grain size of UO ₂ nuclear fuel sintered bodies by the addition of (IV) complexes.

Hereinafter, the present invention will be described in detail.

Uranium (IV) (polyvinylalcohol-uranium, PVA-U (IV)) complex for efficiently treating uranium scrap and uranium ions present in wastewater generated during the process of producing UO ₂ powder (Ii) 2.5 to 7.5 ㅧ 10 ^-3 mole of UO ₂ (or U ₃ O ₈ ) powder was dissolved in 2.5 ml of ⁷ mole nitric acid, and then 97.5 ml of distilled water was added to 200 ml of PVA aqueous solution of 2 to 6 ㅧ 10 ^-2 unit mole U (IV) complex is formed in the range of pH 7 ~ 11 by adding iii) NH ₄ OH solution, and iv) the complex of PVA-U (IV) After filtration, the PVA-U (IV) complex can be prepared by drying in an oven at 60 ° C., and the manufacturing process is shown in FIG.

On the other hand, PVA-U (IV) complexes are expected to form complexes through the following reaction.

PVA is CH-O that is not dissociated in a strong acid atmosphere begins to hydrogen is ionized in the alcohol key of PVA from pH 5.5 part OH dissociates in part as the expression (1) ^ligands is formed.

 The two repeating units of PVA are denoted by HL as follows.

Since uranium nitrate [UO ₂ (NO ₃ ) ₂ ] is mostly present as UO ₂ ⁺⁺ ion in the aqueous solution, the reaction ratio of ionized CH-O ^- and UO ₂ ⁺⁺ ions is as follows.

Thus, the PVA-U (IV) complex has one U (IV) It is considered to be formed with a structure in which four OH groups are bonded around an ion.

As the amount of complex formed increases, the binding of ligand and uranium ion of PVA appears to form a complex with the structure shown in Fig. 2. As the complexation proceeds, the PVA ligand itself is extended but the PVA chain is expected to become significantly compact.

On the other hand, using additives composed of PVA-U (IV) complex UO ₂ nuclear fuel sintered compact is prepared a UO ₂ powder and mixed with the aqueous solution and the PVA-U (IV) complexes prepared from the UN solution of the PVA was added in a ratio of 0.05 ~ 1.0wt% to ⅰ) UO ₂ powder and ⅱ) mixed the UO ₂ powder is a powder preparing step (pre-compressed metal powder, granulation, spheronization and lubricate) the via UO ₂ granulation powder (granule) to create ⅲ) made of UO ₂ the granular powder has a diameter of applying a pressure of 3 ~ 4ton / ㎠ range 10 mm in length and 11 mm in length, and iv) the molded green compact can be produced by sintering in an electric resistance furnace at 1700 to 1750 ° C for 3 to 3.5 hours, and as a PVA-U (IV) complex Using the constituent additive The manufacturing process of UO ₂ fuel sintered body is shown in FIG.

The function of the PVA-U (IV) complex additive will be described in detail as follows.

Ⅰ) Since UO ₂ sintered body is produced by autoclaving UO ₂ powder at a high temperature of about 95% of theoretical density, the density of the pores that have existed after the production by the high temperature and irradiation is increased in the process of burning in the reactor Densification phenomenon occurs. The UO ₂ sintered body was found to be extinguished with small size pores preferentially in the furnace. Small pores are decomposed into vacancies and absorbed and extinguished outside the grain boundaries or sintered bodies. At low temperatures (~ 1300 ° C), the pores are extinguished by irradiation. Therefore, the larger the percentage of small pores, the faster the densification proceeds.

Ii) U-235 atomic fission generates a variety of fission products such as cesium, iodine, krypton, and xenon. The UO ₂ sintered body has a volume increase of about 1.0% per 10,000 MWD / MTU due to the accumulation of such fission products, which is called volume expansion. The volume increase is caused by the accumulation of the fission solid product and the bubbles formed by the gathering of the fission gas products. As the temperature of the nuclear fuel increases, the volume growth rate increases with the degree of combustion. At a temperature higher than 1800 ℃, the volume growth rate decreases because the fission gas is released outside the sintered body by diffusion. Therefore, the volume change (δV / V _o ) of the UO ₂ sintered body is expressed as follows.

Iii) On the other hand, as the degree of combustion increases, the internal pressure of the fuel rod increases due to the increase of the fission gas product, and the corrosive gas such as iodine or cesium causes excessive corrosion of the cladding at the contact point between the UO ₂ sintered body and the cladding. There is a serious problem related to the soundness of the fuel such as accelerating. The gas atoms generated in the particle during the fission are partly dissolved in the UO ₂ matrix due to the action of the fission fragment and the dissolved atoms are diffused into the grain boundary by the concentration gradient of the grain and grain boundaries, Many bubbles formed in the grain boundaries are grown to form pores in the corners of the grain, and the network structure of the tunnels connected with each other is further developed. As a result, the fission gas products are released to the outside of the sintered body. Fission gas emissions increase as the fuel temperature increases, as the grain size decreases, and as the combustion rate of the fuel increases. Since the inner fissile gas is diffused outward from the pores in the UO ₂ sintered body directly to the outer pores, the fission gas is discharged directly out of the sintered body. Therefore, the larger the open pore fraction of the sintered body, the greater the amount of fission gas released.

Ⅳ) Pore structure of UO ₂ sintered body stable in furnace: The pore and pore size distribution of UO ₂ sintered body are the most important factors for densification and the extinction of micropores is faster than extinction of coarse pores among the pores. If spherical pores are stable to densification and there are many initial pores near grain boundaries, densification increases. When the value of the pore diameter is 2 탆 or less, the densification is very large, but when the pore diameter is larger than that, the diameter does not affect the densification. In addition, small pores with a diameter of 2 μm or less are destroyed by fission fragments, and the resulting pores are diffused and trapped by large pores having a diameter of 8 μm or more while moving to grain boundaries, thereby increasing the volume of large pores. Diffusion into the grain boundaries. As the pore size is smaller, the mobility of the pores reaching the equilibrium is inversely proportional to the pore diameter R (M? 1 / R), so that it diffuses into the grain boundaries and is likely to disappear. Also, in the presence of 10 to 30 탆 pores in the UO ₂ sintered body, the fission gas product is absorbed into the pores, thereby suppressing the volume expansion and suppressing the release of the fission gas product to the outside of the sintered body.

Therefore, the PVA-U (IV) complex added to the present invention is added to the UO ₂ powder to make the UO ₂ fuel sintered body, and the PVA is thermally decomposed to form coarse pores of 10 to 30 μm in the UO ₂ sintered body, And the U ₃ O ₈ compound undergoes phase transition after pyrolysis, thereby enabling efficient treatment of uranium scrap.

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 embodiments are merely illustrative of the present invention and that the scope of the present invention is not construed as being limited by these embodiments.

Example 1. Preparation of PVA-U (IV) complex

The PVA-U (IV) complex is prepared by i) dissolving 2.5 ㅧ 10 ^-2 mole of UO ₂ (or U ₃ O ₈ ) powder in 2.5 ml of ⁷ mole nitric acid in 200 ml of PVA aqueous solution of 4 ㅧ 10 ^-2 unit mole (IV) complex was prepared by adding i) NaOH solution to pH 6 ~ 12 to form a PVA-U (IV) complex, and iv) filtering the formed PVA-U (IV) complex And then dried in an oven at 60 ° C to prepare a PVA-U (IV) complex.

Experimental Example 1-1. Confirmation of formation of PVA-U (IV) complex

In order to confirm the formation of the PVA-U (IV) complex prepared in Example 1 above, the PVA and PVA-U (IV) complexes were analyzed by KBr method (with a FT-IR spectrometer (Bruker, VERTEX 80V) KBr: 1/100), the molecular bonding pattern was measured in the range of 4000-400 cm ^-1 . As shown in Fig. 3, PVA showed CH ₂ absorption band at 2940 cm ^-1 and OH absorption band at 3400 cm ^-1 while PVA-U (IV) complex showed CH ₂ Absorption band was almost absent and widely spread over 3600 ~ 3200cm ^-1 area. This is because the absorption band is lost or broadened due to the coordination of uranium ions with PVA.

Experimental Example 1-2. Relative Viscosity Change of PVA-U (IV) Complex

To investigate the formation conditions of the PVA-U (IV) complex prepared in Example 1 above, the relative viscosity of PVA-U (IV) / PVA was measured with a Ostwald viscometer at 25 ° C in a constant temperature bath The ratio was measured. As a result, as shown in Fig. 4, the relative viscosity ratio of the PVA-U (IV) complex in the basic region of the pH 10 portion at pH 8 was small, because there was a change in the aqueous solution due to complex formation. Especially, the lowest value was obtained at pH 9, which is probably due to the increase of complex formation due to the coordination bond between the ligand of PVA chain and U (IV) ion.

Experimental Examples 1-3. U (IV) ion concentration of PVA-U (IV) complex filtrate

In order to investigate the U (IV) ion concentration of the PVA-U (IV) complex filtrate produced in the PVA-U (IV) complex preparation process of Example 1, an induction plasma spectroscopic analyzer (ICP / AES, Perkinelmer Plasma 1000). As a result, as shown in Table 1, it was shown that the concentration of the complex formed by coordination between the ligand of the PVA chain and the U (IV) ion was 0.5 ppm at pH 8, 0.3 ppm at pH 9 and 2 ppm at pH 10 U (IV) in the aqueous solution was decreased.

[Table 1]

Experimental Examples 1-4. Morphology Observation of PVA-U (IV) Complexes

In order to observe the morphology of the PVA and the PVA-U (IV) complex prepared in Example 1 above, PVA and PVA-U (IV) complexes were deposited in gold and then examined using a scanning electron microscope (Tescan, Mira 3 LMU FEG ) Was used to observe the morphology. As a result, as shown in FIG. 5, as compared with PVA, the PVA-U (IV) complex forms a complex between the ligand of the PVA chain and the U (IV) , It is found that the structure formed with the complex is more compact than the PVA.

Experimental Examples 1-5. X-ray diffraction analysis of PVA-U (IV) complex

An X-ray diffractometer (Rigaku, D / Max-2200V) was used to pyrolyze the PVA-U (IV) complex prepared in Example 1 under nitrogen atmosphere at 450 ° C The X-ray used was 40kV, 40mA CuKα, and the diffraction angle (2θ) was in the range of 20-60 ㅀ.

As a result, as shown in Fig. 6, the PVA-U (IV) complex pyrolyzed at 450 ° C shows the amorphous structure in the case of PVA, while the chain of PVA is pyrolyzed at this temperature and the U (IV) It was confirmed that the α-U ₃ O ₈ phase was produced as the main phase.

Example 2. Preparation of PVA-U (IV) complex with UO ₂  Manufacture of nuclear fuel sintered body

A UO ₂ powder ⅰ) Example 1. Preparation of a PVA-U (IV) The UO ₂ powder mixed to the complex was added at a rate of 0.05 ~ 1.0wt%, and produced in ⅱ) UO ₂ powder is mixed powder prepared UO ₂ granule is made through the process (preliminary compaction, granulation, lubrication and lubrication), and iii) the UO ₂ granule powder produced is pressurized in the range of 3.3 ton / The green compacts were sintered for 3 to 3.5 hours at 1750 ℃ in an electric resistance furnace under a hydrogen atmosphere to prepare UO ₂ sintered bodies containing PVA-U (IV) complexes.

Experimental Example 2-1. UO ₂  Density measurement of sintered body

Example 2 divided by the density of the UO ₂ nuclear fuel sintered UO ₂ is obtained by subtracting the weight of the sintered body in a buoyancy medium in the sintered UO ₂ in air using a density measuring instrument (Sartorius Co.) value in distilled water to a value from here And the density (0.9983 g / ㎤) of the buoyancy medium composed of the agepeon surfactant to reduce the surface tension of the water. The calculated sintered density value was divided by the theoretical density of the UO ₂ sintered body (10.96 g / ㎤) Respectively. The results are shown in FIG. 7. As the addition amount of PVA-U (IV) complex increases, the density of the sintered body decreases. When 1.0 wt% of PVA-U (IV) complex is added It was found that the density of the sintered body dropped to about 2.1% TD. It was confirmed that the PVA chain of the PVA-U (IV) complex was pyrolyzed to act as a pore-forming agent.

Experimental Example 2-2. UO ₂  Pore Size Distribution and Grain Size Measurement of Sintered Body

In order to measure the pore size distribution and grain size distribution of the UO ₂ fuel sintered body manufactured in Example 2, the sample was cut with a diamond saw, and the cut specimen was cured using a hardener. rpm for 3 minutes, and polished at 300 rpm for 5 minutes using an Al ₂ O ₃ abrasive, and then the pore size distribution was observed. Thereafter, the polished specimens were heated in a tubular furnace in an oxidizing atmosphere (5 L CO ₂ / min) at about 1300 ° C for 1 hour, cooled to room temperature, and then the grain size was measured. The pore size distribution and grain size were analyzed using an optical microscope (NIKON, ECLIPSE LV150) and a phase analyzer (Media Cybernetics, IMAGE PRO-PLUS).

As a result, as shown in Fig. 8, the pore size distribution of the UO ₂ sintered body to which the pure UO ₂ sintered body and the PVA-U (IV) complex are added shows a pore size of 1 to 2 μm, whereas PVA-U IV) complex was added, it was confirmed that pores having a size of 1 ~ 2 ㎛ and pores having a size of 10 ~ 30 ㎛ appeared together. This is a PVA-U (IV) complexes of UO ₂ sintered body relative to pure water was added UO ₂ sintered body is that the PVA chain of the PVA-U (IV) complex is thermally decomposed hayeotgi forming pores of 10 ~ 30㎛.

On the other hand, the grain size of the UO ₂ sintered body to which the pure UO ₂ sintered body and the PVA-U (IV) complex are added is as shown in FIG. The grain size of the pure sintered body was about 8 ~ 10 ㎛ and the grain size of the sintered body with PVA-U (IV) complex was about 9 ~ 11 ㎛. This indicates that the phase transition to U ₃ O ₈ after thermal decomposition of the PVA-U (IV) complex does not affect the grain size.

Experimental Example 2-3. UO ₂  Observation of sintered density of sintered body

In general, the change in the pore structure occurs even when the sintering by re-condensation occurs to the extent similar to that in the reactor. In order to investigate the thermal stability of the UO ₂ sintered body in the furnace, sintering test was carried out. The sintered body prepared in Example 2 was sintered in a hydrogen gas atmosphere at 1,700 ° C for 24 hours using a sintering test (Centorr SM 60) to determine the degree of sintering. obtaining a difference between the density of the sintered UO ₂ and UO ₂ sintered body after sintering prisoners battle dividing it by the theoretical density (10.96 g / ㎤) was obtained sintered material density change.

The sintered density of the UO ₂ sintered body added with the pure UO ₂ sintered body and the PVA-U (IV) complex after the sintering at 1700 ° C. for 24 hours is shown in Table 2 below.

[Table 2]

As can be seen from Table 2, the pure UO ₂ sintered body case, the sintering density of the sintered material density from 97.71% to 98.26% TD TD is yieoteuna change ratio was 0.55%, PVA-U (IV ) complex of the added sintered UO ₂ The sintered density at 96.35% TD was 96.69% TD and the rate of change was 0.37%. This is because coarse pores having a size of 10 to 30 μm or more exist instead of micropores having a size of 2 μm or less, and thus are stabilized against thermal changes.

Having described specific portions of the present invention in detail, it will be apparent to those skilled in the art that this specific description is only a preferred embodiment and that the scope of the present invention is not limited thereby. It will be obvious. Accordingly, the actual scope of the present invention will be defined by the appended claims and their equivalents.

Claims (8)

(1) preparing an aqueous solution of polyvinylalcohol;
(2) preparing a uranium nitrate (UN) solution by dissolving UO ₂ or U ₃ O ₈ powder in nitric acid and then diluting with UO ₂ or distilled water;
(3) adding the uranium nitrate (UN) solution prepared in the step (2) to the polyvinyl alcohol aqueous solution prepared in the step (1);
(4) adjusting the pH of the solution prepared in the step (3) to 7 to 11 to form a polyvinylalcohol-uranium complex; And
(5) A method for producing a polyvinylalcohol-uranium complex comprising filtering a complex formed by the step (4) and drying the complex.
The method according to claim 1,
Wherein the pH is adjusted to 7 to 11 by adding an NH ₄ OH solution in the step (4). ≪ RTI ID = 0.0 > 11. < / RTI >
A nuclear fuel sintered body produced by adding an additive composed of a polyvinylalcohol-uranium complex to a UO ₂ powder in a uranium dioxide fuel sintered body.
The method of claim 3,
Wherein an additive comprising the polyvinyl alcohol-uranium complex is added in an amount of 0.05 to 1.0 wt.% To UO ₂ powder.
A method for producing a uranium dioxide fuel sintered body,
(1) mixing a UO ₂ powder with an additive comprising a polyvinylalcohol-uranium complex to prepare a mixed powder;
(2) preparing a slug by preliminarily molding the mixed powder produced in the step (1), and then crushing the slug to prepare a granular powder;
(3) compressing and molding the granular powder produced in the step (2) and molding it into a green compact; And
(4) sintering the green compact molded by the step (3) under a hydrogen atmosphere.
6. The method of claim 5,
Wherein the additive comprising polyvinyl alcohol-uranium complex is added in an amount of 0.05 to 1.0 wt% to the UO ₂ powder in the step (1).
6. The method of claim 5,
Wherein the granulated powder is compacted and molded in the step (3) to form a green compact having a diameter of 10 mm and a length of 11 mm.
6. The method of claim 5,
Wherein the sintering is performed for 3 to 4 hours under hydrogen atmosphere at 1700 to 1800 ° C in the step (4).

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