KR20160051113A - Nuclear fuel composite pellets and its fabrication method - Google Patents

Abstract

A nuclear fuel composite pellet of the present invention includes: a granule phase wherein a volume ratio of a non-oxide ceramic nuclear fuel including a nitride or a carbide is equal to or more than 30% and less than 99%, and a volume ratio of an oxide nuclear fuel is equal to or more than 1 and less than 30%; and a matrix phase wherein a volume ratio of the non-oxide ceramic nuclear fuel is equal to or more than 1% and less than 30%, and a volume ratio of the oxide nuclear fuel is equal to or more than 70% and less than 99%. The granule phase is distributed in the matrix phase, and is surrounded by the matrix phase. In the nuclear fuel composite pellet, the density of a fissionable material is high due to a non-oxide ceramic material, as compared to the oxide nuclear fuel, and the nuclear fuel composite pellet has excellent thermal conductivity. Also, the nuclear fuel composite pellet of the present invention can be used as a nuclear fuel of a nuclear power reactor like a light water reactor or a heavy water reactor, since an oxide nuclear fuel layer having strong corrosion resistance against water prevents the non-oxide nuclear fuel from contacting water.

Description

FIELD OF THE INVENTION [0001] The present invention relates to a nuclear fuel composite material sintered body,

The present invention relates to a nuclear fuel composite material sintered body composed of an oxide fuel such as uranium dioxide and a non-oxide ceramic fuel such as a nitride or a carbide, and a method of manufacturing the same. More specifically, It is possible to prevent the non-oxide ceramic fuel material from being in direct contact with water as a coolant by mixing the granules and the base powders so that the volume fraction of the non-oxidized ceramic fuel has a lower matrix structure than the granular phase, The present invention relates to a nuclear fuel sintered body which is stable even to the nuclear fuel rod damage while having high density and thermal conductivity of the nuclear fission element and a method for manufacturing the same.

Oxide fuel such as uranium dioxide is lower in operating margin and economical efficiency due to lower thermal conductivity and cation density than other nuclear fuel materials. However, it has excellent chemical and physical stability in water used as a coolant and is used as a nuclear fuel for commercial power generation such as light water reactor and heavy water reactor.

On the other hand, non-oxide ceramics nuclear fuel such as uranium nitride or uranium carbide have the advantages of high uranium density and high thermal conductivity per unit volume of uranium dioxide. These advantages make it economically feasible to use it as a light-water reactor or heavy-water reactor fuel. However, uranium nitride or uranium carbide easily reacts with water at 350 ° C, losing structural stability. Therefore, when the non-oxide fuel sintered body comes into contact with the cooling water due to the damage of the fuel rod, it becomes powdered and contaminates the entire reactor, so it is not used as a nuclear fuel for a power plant such as a light water reactor or a heavy water reactor.

Recently, efforts are being made to improve the safety and economy of nuclear reactors at the same time. In the case of nuclear fuel, efforts are being made to secure safety by increasing the thermal conductivity and density of fissile material, thereby improving the economical efficiency. (Nucl. Mat., 427 (2012) 209-224) in which nitrite or carbide fuel particles are coated with a triple layer to produce TRISO particles and then the particles are dispersed in a SiC matrix It was proposed. Also, a technique (JP19990202072a2p) in which the surface of the non-oxide fuel particles or the sintered body is coated with a material which does not react with water has been proposed.

The volume of TRISO particles in the sintered body is less than 45% by volume and the amount of fuel that can be charged into the unit volume of sintered body is very small because the volume fraction of the nuclear fuel particles is small even in TRISO. In addition, the coating layer of the particles may interfere with the transfer of heat generated from the fuel to the base, so that the temperature distribution in the particle becomes large. Therefore, it is difficult to maintain the advantages of high uranium density and excellent thermal conductivity of nonoxidized nuclear fuel. In addition, TRISO is produced by forming non-oxidized ceramic particles, forming multiple coating layers on the particles, sintering them in a pressurized environment of high temperature by mixing them with SiC powder, and thus manufacturing process cost is high.

Japanese Unexamined Patent Application Publication (JP19990202072a2p) discloses a nuclear fuel sintered body in which a non-oxide fuel particle having a coating layer is dispersed in an oxide fuel. Since the sintered body uses the oxide as a nuclear fuel rather than the technique utilizing TRISO particles, the amount of uranium loading can be prevented to some extent. It is difficult to avoid reduction of the thermal conductivity by the coating layer and reduction of the uranium loading, and the process cost is also high because the process of manufacturing the particles of the coating layer is added.

In conclusion, safety and economic feasibility of nuclear fuel can be improved by using non-oxide fuel sintered body having high thermal conductivity and high density of nuclear material as nuclear fuel while ensuring safety. Therefore, in order to utilize the non-oxide ceramic fuel as a nuclear fuel for the light water reactor and the heavy water reactor, it is necessary to prevent the non-oxide ceramic fuel material from coming into direct contact with the fuel due to the damage of the nuclear fuel during operation and to develop a sintered body This is desperately needed. However, when other materials than nuclear materials are used as means for preventing contact with water, the amount of uranium is reduced and the economical efficiency is lowered. Therefore, the sintered body is composed of nuclear material and the non- It is urgently necessary to develop a technique for making it possible.

JP19990202072a2p US 2005 / 0286676A1

J. Nucl. Mat., 427 (2012) 209-224

It is an object of the present invention to provide a composite material fuel sintered body having low reactivity with water as a coolant and having excellent density and thermal conductivity of a fission element compared to conventional oxide fuel. Therefore, the volume fraction of the non-oxide ceramic fuel is higher Granular phase and non-oxide ceramic fuel grains are dispersed in an oxide fuel phase without being connected to each other.

Another object of the present invention is to propose a method of manufacturing a nuclear fuel composite material sintered body excellent in compatibility with conventional ceramic processes in which a non-oxide ceramic fuel powder and an oxide fuel powder are simply mixed, formed and sintered.

A fuel composite sinter according to one embodiment of the present invention comprises a granule phase comprising 30 to 99 volume% of a non-oxide ceramic fuel and 1 to 70 volume% of an oxide fuel; And a matrix phase comprising from 1 to 30% by volume of a non-oxide ceramic fuel and from 70 to 99% by volume of an oxide fuel, wherein the granular phase is distributed in the matrix and surrounded by a matrix.

The non-oxide ceramic fuel and oxide fuel may be distributed in the form of grains having a size of 1 nm to 10 mu m in the granular phase and the matrix phase, respectively.

Wherein the non-oxide ceramic fuel comprises any one selected from the group consisting of a mononitride fuel, a monocarbose fuel, and combinations thereof, the oxide fuel being selected from the group consisting of uranium dioxide, plutonium dioxide, thorium dioxide, One can be included.

The ratio of anions to cations in the mono-nitride fuel and monocarbose fuel may be between 0.9 and 1.1.

Wherein the positive element of the single nitrile fuel and the single carbide fuel includes any one selected from the group consisting of uranium, uranium alloy, plutonium, thorium, and combinations thereof, and the uranium alloy is selected from the group consisting of chromium (Cr), zirconium (Zr) And may be an alloy of uranium and any one selected from the group consisting of niobium (Nb), titanium (Ti), silicon (Si), and combinations thereof.

According to another embodiment of the present invention, there is provided a method of manufacturing a sintered fuel composite material, comprising: (a) mixing a non-oxide ceramic fuel powder at 30 to 99 volume% and an oxide fuel powder at 1 to 70 volume% ; (b) mixing the non-oxide ceramic fuel powder at 1 to 30 volume% and the oxide fuel powder at 70 to 99 volume% to prepare a second powder; (c) mixing the first powder and the second powder to form a formed body; And (d) sintering the molded body in an inert gas atmosphere or a vacuum atmosphere.

In the above method, the non-oxide ceramic fuel powder includes any one selected from the group consisting of a mononitride fuel, a monocarbose fuel, and a combination thereof, and the oxide fuel powder includes uranium dioxide, plutonium dioxide, thorium dioxide powder, And a combination of these.

In this method, the ratio of anions to cations in the mono-nitride fuel powder and the mono-carbide fuel powder may be between 0.9 and 1.1.

In the above method, the positive element of the mono-nitride fuel powder and the mono-carbide fuel powder includes any one selected from the group consisting of uranium, uranium alloy, plutonium, thorium, and combinations thereof. The uranium alloy includes chromium, Zirconium, niobium, titanium, silicon, and a combination thereof, and an alloy of uranium.

The step (c) may further comprise a granulation step of preparing a granular phase using the first powder.

The sintering temperature of step (d) may be between 1100 and 2000 ° C.

The inert gas in step (d) may comprise nitrogen gas.

The sintering in the step (d) may be performed by an electric furnace sintering, a hot press sintering, a hot isostatic press (HIP) sintering, a microwave furnace sintering, a spark plasma sintering ), And combinations thereof.

The first powder and the second powder each further include U ₃ Si ₂ as an additive, wherein the U ₃ Si ₂ may be included in an amount of 0.01 to 5% by weight based on the total weight of the molded article.

Wherein the first powder and the second powder each comprise MnO, Y ₂ O ₃ , SiO ₂ , Cr ₂ O ₃ , Al ₂ O ₃ , Zr ₂ O ₃ , TiO ₂ , Nb ₂ O ₅ and combinations thereof And the metal oxide may be contained in an amount of 0.01 to 1% by weight based on the total weight of the molded article.

The present invention provides a sintered fuel composite material and a method of manufacturing the same, and the present invention will be described in detail below.

In the case of nuclear fuel, the density of uranium per unit volume should be high, the better the thermal conductivity, the better, and the reactor should have stability that does not cause an explosive reaction by reacting with water as a coolant. For example, Uranium Nuclear Fuel, which is used as a nuclear fuel for light water reactors and heavy water reactors, has a higher density of uranium per unit volume and higher thermal conductivity than non-oxide ceramic fuel such as uranium nitride or uranium carbide. However, There is a problem in stability. In the case of uranium oxide fuel, it is stable in terms of reactivity with water, but it has a low uranium density and low thermal conductivity, resulting in a considerable decrease in fuel efficiency. Accordingly, in the present invention, it is possible to improve the uranium density and thermal conductivity, which are disadvantages of, for example, uranium oxide fuel, and to prevent the reaction with water, which is a disadvantage of the non-oxide ceramic fuel, by mixing the oxide fuel with uranium oxide, To provide a sintered body of a nuclear fuel composite material.

When the spherical particle powder is randomly dispersed in the matrix, when the volume fraction of the particle powder is low, the individual particle powders are separated from each other and distributed in the matrix as shown in Fig. 1 (a) As shown in FIG. However, as the volume fraction of the particle powder increases, contact between the particles occurs. When the particle fraction increases above a certain fraction, the contacted particles can be connected by a single line as shown in FIG. 1 (b). The volume fraction of the dispersed powder is increased so that the minimum volume at which the particles are connected is called the percolation threshold volume. The volume fraction of the powder that is connected depends on the shape of the powder, but assuming a spherical powder The percolation critical volume is about 30% by volume.

When the fuel composite material sintered body is prepared by simply mixing the non-oxide ceramic fuel powder having a volume fraction of less than 30% by volume and the oxide fuel powder having a volume fraction of not less than 70% by volume, A sintered body in which the non-oxide fuel powder is dispersed in the oxide fuel phase can be obtained. The sintered fuel composite material having such a microstructure prevents a non-oxide ceramic fuel grain from being in contact with water because the oxide fuel material stable in water acts as a protective film. Therefore, the composite fuel having the structure shown in FIG. 1 (a) is likely to be able to prevent corrosion or oxidation caused by water even if it comes into contact with water during use as a light water reactor or a heavy water reactor fuel. However, the fuel of FIG. 1 (a), which is the result of simple mixing of the non-oxide ceramic fuel powder and the oxide fuel powder, is limited in the improvement of the uranium density or thermal conductivity because the volume fraction of the non-oxide ceramic fuel is limited to less than 30% Follow.

Accordingly, if the volume fraction of the non-oxide ceramic fuel is increased to increase the density and the thermal conductivity of the fission element, a non-oxide ceramic fuel wire connected from the surface to the inside of the sintered body is formed as shown in FIG. 1 (b). In this case, when the non - oxide ceramic fuel particles exposed to the surface of the sintered body are in contact with water and are corroded and flowed out, the non - oxide particles on the inner side are exposed to water again,

In order to solve the problems in the nuclear fuel produced by simple mixing of the non-oxide ceramic fuel powder and the oxide fuel powder, the present invention provides a fuel composite sintered body containing a non-oxide ceramic fuel powder and an oxide fuel powder in different volume fractions to provide. Specifically, the sintered fuel composite material of the present invention includes a granular phase and a matrix phase, and both the granular phase and the matrix phase contain both non-oxide ceramic fuel and oxide fuel at the same time. However, on the granular phase, On the base, the fuel composite sintered body is constructed so that the volume fraction of the oxide fuel is higher.

Thus, the fuel composite sinter according to one embodiment of the present invention comprises a granule phase comprising 30 to 99 volume% of a non-oxide ceramic fuel and 1 to 70 volume% of an oxide fuel; And a matrix phase comprising from 1 to 30% by volume of a non-oxide ceramic fuel and from 70 to 99% by volume of an oxide fuel, wherein the granular phase is distributed in the matrix and surrounded by a matrix.

The sintered material of the fuel composite material according to an embodiment of the present invention includes a granule phase and a matrix phase having different volume fractions of the non-oxide ceramic nuclear fuel grains and oxide fuel grains contained therein. Wherein the granular phase is distributed in the matrix and is surrounded by a matrix, and the non-oxide ceramic fuel grains and the oxide fuel grains have sizes ranging from several nanometers to several micrometers in the granular phase and the gaseous phase, respectively, And may be contained in the form of crystal grains.

Wherein the non-oxide ceramic fuel comprises any one selected from the group consisting of a mononitride fuel, a monocarbose fuel, and combinations thereof, the oxide fuel being selected from the group consisting of uranium dioxide, plutonium dioxide, thorium dioxide, One can be included. In addition, the ratio of the positive ion to the negative ion in the mono-nitride fuel and the monocarbose nuclear fuel may be between 0.9 and 1.1, and the cation element of the mono-nitrile fuel and the monocarbose fuel may be uranium, uranium alloy, plutonium, thorium, Wherein the uranium alloy comprises one selected from the group consisting of Cr, Zr, Nb, Ti, Si, and combinations thereof. And an alloy of uranium.

Preferably, the non-oxide ceramic fuel may include any one selected from the group consisting of a nitrile nitride alloy, a uranium carbide alloy, and combinations thereof, and the oxide fuel may include uranium dioxide.

The matrix has non-oxide ceramic fuel grains dispersed randomly in the matrix and surrounded by oxide fuel grains without contact with each other.

The matrix oxide fuel fuel should surround the non-oxide ceramic fuel grains and serve as a barrier to prevent the non-oxide ceramic fuel grains from contacting the surface of the sintered body exposed to the sintered body. For this purpose, it is possible to determine the volume fraction of the non-oxide ceramic fuel and oxide fuel at the base on the basis of the percritical critical volume theory according to the particle shape, and when the spherical crystal grains are present, the volume fraction of the non- Between 1 and 30% by volume and a volume fraction of oxide fuel grains between 70 and 99% by volume.

The volume fraction of the non-oxide ceramic fuel in the granular phase of the fuel composite sintered body may be greater than the volume fraction of the non-oxide ceramic material in the matrix. Thus, the granular phase may comprise between about 30 and about 99 vol% non-oxide ceramic fuel grains and between about 1 and about 70 vol% oxide fuel grains. That is, the granular phase with a high fraction of the non-oxide ceramic fuel material has a shape surrounded by a low-fractionation matrix of the non-oxide ceramic fuel material. The materials contained in the granular phase and the matrix may differ only in volume fraction, and the materials may be uniformly dispersed throughout the granular phase and the matrix in the form of crystal grains, and the size of each of the crystal grains may be several nanometers To several micrometers in size.

Accordingly, the non-oxide ceramic fuel material in the granular phase, which can determine the uranium density and thermal conductivity of the fuel, is provided with a protective film of oxide fuel in contact with water.

In other words, the granular phase with a high fraction of non-oxide ceramic fuel fuels is surrounded by the matrix. At this time, the fraction of the non-oxide ceramic fuel included in the matrix is less than 30% The non-oxide ceramic fuel particles are surrounded by oxide fuel. Relatively, the granular phase has a shape similar to that of Fig. 1 (b). Therefore, in the composite sintered body including the matrix of such a structure, uranium oxide powder in the granular phase is prevented from coming into contact with water, and even if oxidation of the partially exposed non-oxidized ceramic fuel particles occurs in the matrix, The integrity of the sintered body can be maintained.

Also, in terms of efficiency as a nuclear fuel, even though the non-oxide ceramic material is contained in both the matrix and the granular phase, even if the fraction of the non-oxide ceramic material on the matrix is low, It is possible to greatly improve the density and the thermal conductivity of the fission element.

The granular phase and the matrix phase contained in the nuclear fuel sintered body are not particularly determined in terms of their volume ratio. However, since the distance between the granular phase and the granular phase should be larger than the size of the non-oxide ceramic particles, since the granular phase completely covers the granular phase, it is preferable that the granular phase volume ratio is not less than 10% and not more than 80%.

The non-oxide ceramic fuel material is preferably uranium mononitride or uranium monocarbide. Uranium is an element composed of plutonium, thorium, or a combination of uranium, plutonium, thorium, or a uranium alloy . ≪ / RTI > The uranium alloy may be one in which the fuel material is replaced by a small amount of other cations in order to improve the corrosion resistance of the non-oxide ceramic fuel material. Examples thereof include Cr, Zr, Nb, , Silicon (Si), or a combination thereof may be used in place of uranium as a fission element. The composition of these alloys can be equally applied to plutonium and thorium. When a part of the additive element is replaced with the additive element, the density and thermal conductivity per unit volume of uranium may be lowered, but it is possible to lower the rate of corrosion of uranium nitride or monocarborylium in water. Therefore, whether the additive element is substituted or not can be selected in consideration of the performance of the fuel and the corrosion stability.

According to another embodiment of the present invention, there is provided a method of manufacturing a sintered fuel composite material, comprising: (a) mixing a non-oxide ceramic fuel powder at 30 to 99 volume% and an oxide fuel powder at 1 to 70 volume% ; (b) mixing the non-oxide ceramic fuel powder at 1 to 30 volume% and the oxide fuel powder at 70 to 99 volume% to prepare a second powder; (c) mixing the first powder and the second powder to form a formed body; And (d) sintering the molded body in an inert gas atmosphere or a vacuum atmosphere.

In the above method, the non-oxide ceramic fuel powder includes any one selected from the group consisting of a mononitride fuel, a monocarbose fuel, and a combination thereof, and the oxide fuel powder includes uranium dioxide, plutonium dioxide, thorium dioxide powder, And a combination of these.

In this method, the ratio of anions to cations in the mono-nitride fuel powder and the mono-carbide fuel powder may be between 0.9 and 1.1.

In the above method, the positive element of the mono-nitride fuel powder and the mono-carbide fuel powder includes any one selected from the group consisting of uranium, uranium alloy, plutonium, thorium, and combinations thereof. The uranium alloy includes chromium, Zirconium, niobium, titanium, silicon, and a combination thereof, and an alloy of uranium. The steps (a) and (b) are the steps of preparing the first powder and the second powder, respectively, and powder can be produced by the same method, and can be independently performed independently of the order of each other.

The first powder may comprise from 30 to 99% by volume of the non-oxide ceramic material and from 1 to 70% by volume of an oxide fuel material such as uranium oxide, wherein the second powder comprises from 1 to 30% by volume of the non-oxide ceramic fuel powder, Fuel material, such as 70 to 99 vol% uranium dioxide powder.

The first powder and the second powder may include a non-oxide ceramic fuel powder and a uranium dioxide powder, and may have the same composition ratio and composition, but the volume fraction of the non-oxide ceramic fuel powder and the uranium dioxide powder Or may be a different powder having a different composition ratio. In other words, considering the efficiency of the nuclear fuel and the reactivity with water, the second powder as the gaseous phase contains a low fraction of the non-oxide ceramic fuel powder so as to limit the contact of the non-oxide ceramic fuel grain with water, The first powder that is different may include a high fraction of the non-oxide ceramic fuel powder so that the density and thermal conductivity of the fission element can be increased.

The non-oxide ceramic fuel powders of the steps (a) and (b) may be prepared by applying nitriding or carbonizing treatment of uranium metal or uranium alloy, uranium mononitride, uranium mononitride or uranium mononitride alloy, can do.

Uranium alloy, it is preferable to use at least one metal element selected from the group consisting of chromium (Cr), zirconium (Zr), niobium (Nb), titanium (Ti), silicon (Si) A configured alloy may be applied.

Nitrogen and / or ammonia gas may be used for the nitriding treatment, and carbon and / or methane gas may be used for the carbonization treatment, but the present invention is not limited thereto.

In order to promote the pulverization of uranium or uranium alloys and promote the nitridation reaction or the carbonization reaction, the hydrothermal treatment is performed in the hydrogen gas by the pretreatment process,

When the raw material powder is an oxide, nitriding powder and carbonized powder can be prepared by a thermal carbon reduction reaction method.

The steps (a) and (b) may be performed by mixing the non-oxide ceramic fuel powders such as uranium mononitride or uranium mononitride and the uranium dioxide powder in the predetermined volume ratio as described above.

The step (c) may further comprise a granulation step of preparing a granular phase using the first powder.

The step (c) may be a step of mixing the two powders prepared in steps (a) and (b), wherein the first powder is prepared into granules and then mixed before mixing. The granulating step may be a step of preparing a general granule such that the powder is absorbed in a high temperature and high humidity environment, followed by drying and pulverizing again. Through this process, the first powder can be produced into a granular form . The molding process may also be a molding process which is generally performed. In this case, the granulated first powder and the powdered second powder may be injected into a specific mold and subjected to predetermined heat and pressure to form a molded body.

The step (d) may be a step of sintering the molded body under an inert gas atmosphere or a vacuum atmosphere. The inert gas may be a gas such as helium, argon, etc., and may include nitrogen gas, and the vacuum pressure range may be about 10 ^-3 torr or less.

The temperature for sintering in step (d) may be about 1100 to 2000 ° C. If the sintering temperature is lower than 1100 ° C, there is a fear that the powder is not pressed properly, and there is a possibility that a void or the like is formed on the matrix phase surrounding the granular phase. If the sintering temperature exceeds 2000 ° C, The ceramic fuel powder may be reduced to a metal phase having a low melting temperature.

The sintering in the step (d) may be performed by a general sintering method, for example, electric furnace sintering, hot press sintering, hot isostatic press (HIP) sintering, a microwave furnace sintering, a spark plasma sintering (SPS), or a combination thereof.

The sintering aid may be added to the first powder and the second powder simultaneously or selectively, and the sintering aid may be one selected from the group consisting of U ₃ Si ₂ , SiO ₂ , MnO, Y ₂ O ₃ , Cr ₂ O ₃ , Al ₂ O ₃ , Zr ₂ O ₃ , TiO ₂ , Nb ₂ O ₅ or a combination thereof. Such a sintering aid promotes the densification behavior of the sintered body, thereby lowering the sintering temperature or increasing the density of the sintered body.

When U ₃ Si ₂ is added as the sintering aid, 0.01 to 30% by weight may be added based on the total weight of the formed body, and it is appropriate to adjust the sintering temperature of step (d) to about 1660 to 1750 ° C .

Further, as the sintering aid SiO of _{2, MnO, Y 2 O 3} , Cr 2 O 3, Al 2 O 3, Zr 2 O 3, TiO 2, Nb 2 O 5 and at least one selected from the group consisting of When the metal oxide is added, 0.01 to 1% by weight may be added based on the total weight of the molded article, and at this time, it may be appropriate to adjust the sintering temperature of step (d) to about 1400 to 2000 ° C.

Since the content of the sintering aid does not affect the volume ratio of the non-oxide ceramic fuel and the uranium dioxide contained in the sintered fuel composite material, the sintering aid can sufficiently perform the role of the sintering auxiliary agent. Density and thermal conductivity can be reduced. The sintering temperature according to the additive reflects the characteristics of each additive. The sintering temperature is not limited to the ideal temperature range and is not limited to the temperature range. Regardless of the additive, Lt; / RTI >

The present invention relates to a nuclear fuel sintered body which consists of a non-oxide fuel phase and an oxide fuel phase, excluding a coating layer or a coating layer which reduces uranium density and lowers the thermal conductivity. In the sintered body of the present invention, the oxide fueled phase functions as a fuel and serves as a protective film for preventing direct contact of the non-oxide ceramic fueled phase with water. Such a fuel can significantly reduce the decrease in the amount of uranium charge and the decrease in thermal conductivity compared to a coating / coating layer that does not contain a nuclear material or a nuclear material that uses a non-nuclear material as a base material. It is also possible to prevent unexpected reaction between the coating layer or the coating layer and the inner surface of the cladding tube, and to avoid degradation of the dimensional stability of the sintered body due to the coating layer having a melting temperature lower than that of the ceramic fuel such as metal. In addition, since the sintered composite fuel material of the present invention is manufactured by simply mixing and sintering the non-oxide ceramic fuel powder and the oxide fuel powder in the process, it is almost the same as the conventional process and there is no additional burden on the manufacturing process of the nuclear fuel sintering material, none.

The sintered fuel composite material of the present invention has significantly improved density and thermal conductivity of the fission element compared to the uranium dioxide fuel sintered body of the present invention, greatly improving the operation margin, PCI characteristics, and accident resistance of the fuel rod, The cycle economical efficiency can be greatly increased. Uranium dioxide fuel layer serves as a protective film, so that when the nuclear fuel rod is damaged, the nitrite or carbide nuclear fuel grain is limited in direct contact with cooling water, so that it has high resistance to corrosion in water. Therefore, uranium dioxide can be used instead of uranium dioxide as a light- have. In addition, since nitriding or carbide particles are coated with an anti-corrosive coating layer, nitrided or carbide powder and uranium dioxide powder are mixed to produce a sintered body through a conventional ceramic manufacturing process. Can be minimized.

FIG. 1A is a schematic view of a sintered body microstructure in which spherical particle powders are randomly distributed in a matrix when a volume fraction is smaller than a percolation critical volume.
FIG. 1B is a schematic view of a sintered body microstructure in which spherical particle powders are randomly distributed in a matrix when volume fraction is larger than a percolated critical volume.
2 is an SEM image of the sintered body microstructure according to an embodiment of the present invention.
FIG. 3 is an image obtained by enlarging and observing area A according to FIG.
4 is an enlarged image of the region B shown in FIG.
5 is an SEM image of the microstructure of the sintered body produced according to Comparative Example 1. Fig.
6 is an SEM image of the microstructure of the sintered body produced according to Comparative Example 2. Fig.
FIG. 7 is a graph showing the thermal diffusivity variation of the sintered bodies manufactured according to Examples and Comparative Examples according to the temperature, together with the thermal diffusivity of UO ₂ .
FIG. 8 is a photograph of the sintered bodies manufactured according to the examples and the comparative examples to compare changes in sintered body shape after heat treatment in a 350 ° C. steam atmosphere for 12 hours.

Hereinafter, embodiments of the present invention will be described in detail with reference to the accompanying drawings so that those skilled in the art can easily carry out the present invention. The present invention may, however, be embodied in many different forms and should not be construed as limited to the embodiments set forth herein.

Manufacturing example  1: Manufacture of sintered body of fuel composite material

1) preparation of the first powder and Granulation

The spherical uranium powder prepared by the centrifugal atomizer method was heat treated in a hydrogen gas atmosphere at 170 ° C and then in a nitrogen gas atmosphere at 600 ° C. The powder was then heat-treated again in an inert atmosphere at 1400 ° C to produce unaltered ceramic fueled uranium nitride powder. There is a small amount of uranium dioxide in the prepared mononitride powder. As the oxide fuel powder, a uranium dioxide powder was prepared.

The monolith nitrile powder and the uranium dioxide powder were mixed so that the volume fraction of the uranium nitride monolith was 70% by volume, a small amount of organic material for coagulation was mixed, and then the alcohol was pulverized and mixed with the medium. The wet pulverized powder was dried in a glove box at room temperature and then pulverized to prepare a granulated first powder, that is, a granular phase.

2) Preparation of second powder

The same powder as that of the first powder prepared in 1) above and the uranium dioxide powder and the uranium dioxide powder were mixed so that the volume fraction of uranium nitride monofilaments was 15% by volume to prepare a second powder.

3) Shaped body  Formation and Sintering

The granular phase of the first powder prepared in 1) above and the second powder prepared in 2) are mixed in a tubular mixer so that the volume fraction of the first powder is 65% by volume to prepare a mixed powder. Mixed powders were prepared by sintering by hot press method. The mixed powder was injected into the mold and pressurized at a pressure of about 40 MPa at a temperature of 1590 ° C to produce a sintered fuel composite material matrix having a matrix formed thereon.

The density of the sintered body was measured using the Archimedes method. After measuring the density, the cross section of the sintered body was mirror polished and the microstructure was observed through a scanning electron microscope. The phases constituting the sintered body were identified by X-ray diffraction and the volume ratios of uranium nitride and uranium dioxide constituting the sintered body were calculated by comparing the diffraction linewidths of the respective phases. The thermal diffusivity of the sintered body was measured by laser flash method using a disc - shaped sintered body with a thickness of about 3 mm. In order to observe the oxidative behavior of the sintered body in the steam atmosphere, the shape change of the sintered body was observed after heat treatment at 350 ℃ for 12 hours in water vapor atmosphere.

compare Manufacturing example  One

Using the same monovalent nitrifluoride powder and uranium dioxide powder as in Preparation Example 1, they were simply mixed as they were in Preparation Example 1, except that the granules were prepared at different volume fractions. The powder was prepared in Preparation Example 1 To obtain a sintered body. At this time, a sintered body having a volume fraction of uranium nitride monolith powder in the sintered body of about 22% by volume was prepared. The density, the volume fraction of uranium nitride, the thermal diffusivity, and the oxidative behavior of the steam atmosphere were observed for the sintered body through the same procedure as in Production Example 1.

compare Manufacturing example  2

Using the same monovalent nitrifluoride powder and uranium dioxide powder as in Preparation Example 1, they were simply mixed as they were in Preparation Example 1, except that the granules were prepared at different volume fractions. The powder was prepared in Preparation Example 1 To obtain a sintered body. At this time, a sintered body having a volume fraction of uranium nitride monolith powder in the sintered body was about 7 vol%. The density, the volume fraction of uranium nitride, the thermal diffusivity, and the oxidative behavior of the steam atmosphere were observed for the sintered body through the same procedure as in Production Example 1.

FIG. 2 shows the result of observation of the sintered body manufactured in Production Example 1 through a scanning electron microscope. It can be confirmed that the structure in which the granular phase A region is surrounded by the base region B is formed.

Figs. 3 and 4 show enlargement and observation results of regions A and B, respectively. Referring to FIG. 3, it can be seen that bright granular U-shaped crystal grains occupy a considerable portion of the granular A region as compared with dark U-shaped granular crystals. Referring to FIG. 4, As a result of enlargement of the area, it can be confirmed that dark uranium dioxide occupies a considerable portion in contrast to the granular image in Fig.

Fig. 5 shows the microstructure of the sintered body produced by Comparative Example 1. Fig. It can be seen that unlike the sintered body of Production Example 1, the uranium nitride grains are uniformly distributed in the entire sintered body.

6 shows the microstructure of the sintered body produced by Comparative Example 2. Fig. It can be seen that a very thick uranium dioxide layer surrounds the uranium nitride monocrystalline because the volume ratio of the monocrystalline uranium nitride grains is small.

The density of the sintered product of Production Example 1 was 11.8 g / cm ^{3, which} was about 8% higher than that of UO _2. The volume fraction of uranium nitride in the sintered body measured by XRD was about 25% by volume. This translates into an increase in uranium density of about 10% in terms of uranium density. In comparison, the sintered product density of Comparative Example 1 is 11.7 g / cm ³ and the volume fraction of uranium nitride is about 22% by volume. It has a value similar to the density of the sintered product and the volume fraction of uranium nitride monodispersed in Production Example 1. The density of the sintered body of Comparative Example 2 is 10.08 g / cm ³ and the volume fraction of uranium nitride is about 5% by volume.

FIG. 7 shows the thermal diffusivity of the sintered bodies manufactured in Production Example 1 and Comparative Example 1 together with the thermal diffusivity of the UO ₂ sintered body. Production Example 1 and Comparative Example 1 The thermal diffusivity of the sintered bodies greatly increases in comparison with the thermal diffusivity of UO ₂ . In the case of Production Example 1, the thermal diffusivity is increased by 40 to 50% compared to UO ₂ , and in the case of Comparative Example 1, it is increased by 30 to 40%. Despite the sintered bodies of Production Example 1 and Comparative Example 1 having a similar density and a uniform uranium nitride volume fraction, the sintered body of Production Example 1 has a high thermal diffusivity. This is because the volume fraction of uranium nitride in the sintered product of Production Example 1 is higher than that of Comparative Example 1 and the area of the interface of the hetero-crystal grain between uranium nitride monocrystalline grains and uranium dioxide crystal grains is reduced in the sintered product of Production Example 1 . The sintered body of Comparative Example 1 has a relatively large area of the interface of the hetero-grains relative to the sintered body of Production Example 1 because most of the monocrystalline uranium nitride grains are surrounded by uranium dioxide . This fact confirms that the microstructure of the fuel composite sintered body of the present invention is advantageous for enhancing thermal conductivity. The sintered body of Comparative Example 2 showed a 0 to 10% increase in thermal diffusivity compared to the UO ₂ sintered body.

FIG. 8 is a graph comparing the shapes of sintered bodies obtained by heat-treating the three kinds of sintered bodies manufactured in Production Example 1, Comparative Example 1 and Comparative Example 2 in a steam atmosphere at 350 ° C. for 12 hours. The sintered body of Production Example 1 maintained the shape of the sintered body soundly, except that some corners were broken even after the oxidation test. The sintered body of Comparative Example 1 was much less oxidized and fractured than the sintered body of Production Example 1 in spite of the small volume ratio of uranium nitride, and the initial shape of the sintered body was almost lost. In the sintered body of the present invention, it can be confirmed that even if the unidirectionally-grown uranium nitride grains are partially oxidized in the surface of the sintered body, the uranium dioxide base layer covers the unidirefined nitrided nitrides, thereby fulfilling the role of a protective film for preventing further oxidation. However, in the sintered body of Comparative Example 1, when oxidation of monolith nitrides exposed on the surface occurs, the oxidation of the monolith uranium nitrides connected to each other occurs consecutively, so that the sintered body shape can not be maintained. Therefore, even if a composite sintered body in which uranium nitride monoliths with the same volume fraction is mixed with uranium dioxide is used, the sintered body according to the present invention can be used as a light water reactor or a heavy water reactor fuel using a coolant as water, but the sintered body according to Comparative Example 1 can not be used as a nuclear fuel .

In the case of the sintered body of Comparative Example 2, the shape of the sintered body is relatively well maintained even after the steam atmosphere heat treatment as in the sintered body of Production Example 1. This is because the volume fraction of uranium nitride is low, so that in a sintered body having a small volume of uranium nitride distributed in the uranium dioxide, the uranium dioxide layer acts as a protective film to prevent the oxidation of uranium nitride.

While the present invention has been particularly shown and described with reference to exemplary embodiments thereof, it is to be understood that the invention is not limited to the disclosed exemplary embodiments, Of the right.

Claims (15)

A granule phase comprising 30 to 99% by volume of a non-oxide ceramic fuel and 1 to 70% by volume of an oxide fuel fraction; And
A matrix phase comprising from 1 to 30 vol% of the non-oxide ceramic fuel and from 70 to 99 vol% of the oxide fuel,
Wherein the granular phase is distributed in the matrix phase and is surrounded by a matrix. The method according to claim 1,
Wherein the non-oxide ceramic fuel and the oxide fuel are distributed in the form of grains having a size of 1 nm to 10 占 퐉 in the granular phase and in the matrix phase, respectively. The method according to claim 1,
Wherein the non-oxide ceramic fuel includes any one selected from the group consisting of a mono-nitride fuel, a mono-carbide fuel, and a combination thereof,
Wherein the oxide fuel comprises any one selected from the group consisting of uranium dioxide, plutonium dioxide, thorium dioxide, and combinations thereof. The method of claim 3,
Wherein the ratio of anions to cations in said monolithic nuclear fuel and monocarbose fuel is between 0.9 and 1.1. The method of claim 3,
The cation element of the single nitrile fuel and the single carbide nuclear fuel includes any one selected from the group consisting of uranium, uranium alloy, plutonium, thorium, and combinations thereof.
Wherein the uranium alloy is an alloy of uranium with any one selected from the group consisting of chromium (Cr), zirconium (Zr), niobium (Nb), titanium (Ti), silicon . (a) mixing the non-oxide ceramic fuel powder at 30 to 99% by volume and the oxide fuel powder at 1 to 70% by volume to prepare a first powder;
(b) mixing the non-oxide ceramic fuel powder at 1 to 30 volume% and the oxide fuel powder at 70 to 99 volume% to prepare a second powder;
(c) mixing the first powder and the second powder to form a formed body; And
(d) sintering the molded body in an inert gas atmosphere or a vacuum atmosphere
Wherein the sintered body is formed of a material having a high thermal conductivity. The method according to claim 6,
Wherein the non-oxide ceramic fuel powder comprises any one selected from the group consisting of a mononitride fuel, a monocarbose fuel powder, and a combination thereof,
Wherein the oxide fuel powder comprises any one selected from the group consisting of uranium dioxide, plutonium dioxide, thorium dioxide powder, and combinations thereof. 8. The method of claim 7,
Wherein the ratio of the positive ion to the negative ion in the mono-nitride fuel powder and the mono-carbide fuel powder is between 0.9 and 1.1. 8. The method of claim 7,
Wherein the positive element of the mono-nitride fuel powder and the mono-carbide fuel powder includes any one selected from the group consisting of uranium, uranium alloy, plutonium, thorium, and combinations thereof,
Wherein the uranium alloy is an alloy of uranium with any one selected from the group consisting of chromium, zirconium, niobium, titanium, silicon, and combinations thereof. The method according to claim 6,
Wherein the step (c) further comprises a granulation step of producing a granular phase using the first powder. The method according to claim 6,
Wherein the sintering temperature of step (d) is 1100 to 2000 ° C. The method according to claim 6,
Wherein the inert gas of step (d) comprises nitrogen gas. The method according to claim 6,
The sintering in the step (d) may be performed by an electric furnace sintering, a hot press sintering, a hot isostatic press (HIP) sintering, a microwave furnace sintering, a spark plasma sintering ), And a combination thereof. ≪ RTI ID = 0.0 > 11. < / RTI > The method according to claim 6,
Wherein the first powder and the second powder each further comprise U3Si2 as an additive,
Wherein the U3Si2 is contained in an amount of 0.01 to 5 wt% based on the total weight of the molded body. The method according to claim 6,
Wherein the first powder and the second powder each further comprise any one metal oxide selected from the group consisting of MnO, SiO2 Y2O3, Cr2O3, Al2O3, Zr2O3, TiO2, Nb2O5, and combinations thereof,
Wherein the metal oxide is contained in an amount of 0.01 to 1% by weight based on the total weight of the molded article.

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