Neutron-absorbing composite material and storage container for spent nuclear fuel including same

Abstract

Disclosed is a neutron-absorbing composite material. The neutron-absorbing composite material comprises: an aluminum substrate; and reinforcing material powder dispersed in the aluminum substrate and having a boron carbide core and a gadolinium oxide shell covering the surface of the boron carbide core.

Description

Neutron-absorbing composite material and spent nuclear fuel storage container including the same

[0001] The present invention relates to a neutron-absorbing composite material and a spent nuclear fuel storage container including the same.

[0002] Spent nuclear fuel is nuclear waste that emits radioactivity and heat after being used as fuel in a nuclear reactor and must be safely managed in that state. As nuclear fuel burns in a reactor, radioactive materials are generated by nuclear reactions, and these materials emit various forms of radiation, including neutrons and gamma rays. To safely store spent nuclear fuel, storage containers capable of absorbing neutrons are required. Previously, a method was applied to increase the neutron absorption rate by adding gadolinium oxide (Gd2O3) to aluminum-boron carbide (B4C) composites as neutron absorbers. However, this method has the disadvantage that gadolinium oxide does not disperse well within the aluminum-boron carbide composite, leading to magnetic shielding and channeling.

[0003] One objective of the present invention is to provide a neutron-absorbing composite material having excellent neutron absorption ability by improving the magnetic deactivation and shielding ability of gadolinium oxide.

[0004] Another objective of the present invention is to provide a method for manufacturing the neutron-absorbing composite material with increased productivity and processability.

[0005] Another objective of the present invention is to provide a spent nuclear fuel storage container comprising the neutron-absorbing composite material.

[0006] To achieve the above objective, the present invention provides a neutron-absorbing composite material comprising: an aluminum substrate; and reinforcing particles dispersed within the aluminum substrate, the reinforcing particles comprising a boron carbide core and a gadolinium oxide shell covering the surface of the boron carbide core.

[0007] In addition, the present invention provides a method for manufacturing a neutron-absorbing composite material comprising the steps of: manufacturing a reinforcing powder coated with gadolinium oxide on the surface of boron carbide; mixing the reinforcing powder and aluminum powder to manufacture a mixed powder; and processing the mixed powder to manufacture a composite material in which the reinforcing powder is dispersed within an aluminum substrate.

[0008] In addition, the present invention provides a spent nuclear fuel storage container comprising: a basket capable of receiving spent nuclear fuel; a cask disposed to surround the outside of the container; a neutron shielding material disposed to surround the outside of the cask; and an outer casing disposed to surround the outside of the neutron shielding material, wherein the neutron shielding material comprises the neutron-absorbing composite material.

[0009] According to the present invention, the absorbent material of the present invention is introduced into a metal substrate by coating Gd2O3 on the surface of a reinforcing material, thereby reducing the amount of reinforcing material introduced, and Gd2O3 is distributed in a thin thickness through the coating, which can reduce magnetic shielding, and the channeling phenomenon in which neutrons free pass can be reduced by increasing the specific surface area.

[0010] FIG. 1 is a flowchart illustrating a method for manufacturing a neutron-absorbing composite material according to one embodiment of the present invention.

[0011] FIG. 2 is a schematic diagram of a neutron-absorbing composite material according to one embodiment of the present invention.

[0012] Figure 3 is a graph showing the macroscopic cross-sectional areas of Example 1 and Comparative Examples 1 to 6.

[0013] Figure 4 shows the thermal conductivity measurement results of Example 1, Comparative Example 3, and Comparative Example 6.

[0014] Hereinafter, embodiments of the present invention will be described in detail with reference to the attached drawings. Since the present invention is susceptible to various modifications and may take various forms, specific embodiments are illustrated in the drawings and described in detail in the text. However, this is not intended to limit the present invention to the specific disclosed forms, and it should be understood that the invention includes all modifications, equivalents, and substitutions that fall within the spirit and scope of the invention. Similar reference numerals have been used for similar components in the description of each drawing. In the attached drawings, the dimensions of the structures are shown enlarged compared to the actual dimensions for the clarity of the present invention.

[0015] Terms such as "first," "second," etc., may be used to describe various components, but said components should not be limited by said terms. These terms are used solely for the purpose of distinguishing one component from another. For example, without departing from the scope of the present invention, the first component may be named the second component, and similarly, the second component may be named the first component.

[0016] The terms used in this application are used merely to describe specific embodiments and are not intended to limit the invention. The singular expression includes the plural expression unless the context clearly indicates otherwise. In this application, terms such as "comprising" or "having" are intended to specify the existence of the features, numbers, steps, actions, components, parts, or combinations thereof described in the specification, and should be understood as not precluding the existence or addition of one or more other features, numbers, steps, actions, components, parts, or combinations thereof.

[0017] Meanwhile, unless otherwise defined, all terms used herein, including technical or scientific terms, have the same meaning as generally understood by those skilled in the art to which the present invention pertains. Terms such as those defined in commonly used dictionaries should be interpreted as having a meaning consistent with their meaning in the context of the relevant technology, and should not be interpreted in an ideal or overly formal sense unless explicitly defined in this application.

[0018] FIG. 1 is a flowchart illustrating a method for manufacturing a neutron-absorbing composite material according to one embodiment of the present invention.

[0019] Referring to FIG. 1, a method for manufacturing a neutron-absorbing composite material according to one embodiment of the present invention may include the step of manufacturing a reinforcing powder coated with gadolinium oxide on the surface of boron carbide (S110); the step of manufacturing a mixed powder by mixing the reinforcing powder and aluminum powder (S120); and the step of manufacturing a composite material in which the reinforcing powder is dispersed inside an aluminum substrate by processing the mixed powder (S130).

[0020] In the step (S110) of manufacturing a reinforcing powder coated with gadolinium oxide on the surface of the boron carbide, no special restrictions are placed on the method of coating the gadolinium oxide on the surface of the boron carbide. The reinforcing powder can be manufactured by immersing the boron carbide powder in a solution in which a gadolinium oxide precursor is dissolved, then reducing the gadolinium from the gadolinium oxide precursor on the surface of the boron carbide powder, and subsequently heat-treating it.

[0021] In one embodiment, the reinforcing powder may comprise a boron carbide core and a gadolinium oxide shell covering the surface of the boron carbide core. The reinforcing powder may reduce the content of the boron carbide core by means of the gadolinium oxide shell, thereby increasing productivity and processability.

[0022] In the step (S120) of preparing a mixed powder by mixing the reinforcing powder and the aluminum powder, no special restrictions are placed on the method of mixing the reinforcing powder and the aluminum powder. In one embodiment, the reinforcing powder and the aluminum powder may be mixed by ball-milling after being introduced into a milling container.

[0023] In the step (S130) of manufacturing a composite material in which the reinforcing powder is dispersed within an aluminum substrate by processing the above-mentioned mixed powder, the processing may be casting or powder metallurgy, but is not limited thereto.

[0024] In one embodiment, a reinforcing material including the gadolinium oxide shell can be dispersed within the aluminum substrate through the processing. As the gadolinium oxide is uniformly dispersed within the aluminum substrate in a shell form, the magnetic shielding and channeling phenomena of the gadolinium oxide can be mitigated.

[0025] FIG. 2 is a schematic diagram of a neutron-absorbing composite material according to one embodiment of the present invention.

[0026] Referring to FIG. 2, a neutron-absorbing composite material according to one embodiment of the present invention can be manufactured by the above manufacturing method and may include an aluminum substrate; and reinforcing particles dispersed within the aluminum substrate and having a boron carbide core and a gadolinium oxide shell covering the surface of the boron carbide core.

[0027] In one embodiment, the neutron-absorbing composite material contains gadolinium oxide, which has a neutron absorption rate 40 times higher than that of boron carbon, thereby improving the neutron absorption rate compared to the existing Al-B4C composite material. In addition, the magnetic shielding and tunneling phenomena of gadolinium oxide can be overcome by coating the gadolinium oxide onto the boron carbon and dispersing it on an aluminum substrate.

[0028] The thickness of the aluminum substrate may be 0.01 to 5 cm, 0.1 to 30 mm, or 0.1 to 20 mm. If the thickness of the aluminum substrate is less than about 0.1 mm, the distance through which neutrons pass is short, and the neutron absorption capacity may be reduced. If the thickness of the aluminum substrate exceeds about 20 mm, the cost may increase, and it may be disadvantageous for optimizing the design of the spent nuclear fuel storage container.

[0029] The average size of the boron carbide core may be 0.1 to 500 μm or 0.1 to 300 μm. If the average size of the boron carbide core is less than about 0.1 μm, the boron carbides may aggregate as the specific surface area of ​​the boron carbides increases. If the average size of the boron carbide core exceeds 300 μm, the neutron-absorbing composite material may be vulnerable to impact or breakage.

[0030] The average thickness of the gadolinium oxide shell may be 0.01 to 20 μm or 0.01 to 10 μm. If the average thickness of the gadolinium oxide shell is less than 0.01 μm, neutrons may pass through without encountering the gadolinium oxide. If the average thickness of the gadolinium oxide shell exceeds 10 μm, the neutron absorption capacity may be reduced due to magnetic shielding of the gadolinium oxide.

[0031] The content of the reinforcing particles may be 50 vol.% or less or 40 to 50 vol%, and the content of the gadolinium oxide may be 2 vol.% or less or 0.1 to 2 vol.%.

[0032] A spent nuclear fuel storage container according to one embodiment of the present invention comprises: a basket capable of receiving spent nuclear fuel; a cask disposed to surround the outside of the container; a neutron shielding material disposed to surround the outside of the cask; and an outer casing disposed to surround the outside of the neutron shielding material, wherein the neutron shielding material may comprise the neutron-absorbing composite material.

[0033] In one embodiment, the spent nuclear fuel storage container includes the neutron-absorbing composite material, so that it can absorb more than 90% of the neutrons generated in the spent nuclear fuel and effectively release the heat generated in the spent nuclear fuel by increasing the thermal conductivity.

[0034] Hereinafter, the present invention will be described in detail with reference to examples to aid in understanding. However, the following examples are merely illustrative of the content of the present invention and the scope of the present invention is not limited to the following examples. The examples of the present invention are provided to more completely explain the present invention to those with average knowledge in the art.

[0035] <Example 1>

[0036] Boron carbide was immersed in a solution in which gadolinium oxide was dissolved, and the immersed boron carbide was heat-treated to produce a reinforcing powder coated with gadolinium oxide on the surface of the boron carbide. The reinforcing powder and aluminum powder were prepared such that the boron carbide content was 15 vol%. The reinforcing powder and aluminum powder were ball-milled and mixed. The mixed powder was processed by methods such as casting and powder metallurgy to produce a neutron-absorbing composite material. The neutron-absorbing composite material of Example 1 was analyzed using an inductively coupled plasma (ICP) analyzer and confirmed to have a gadolinium oxide content of approximately 0.355 vol%.

[0037] <Comparative Examples 1 to 6>

[0038] A neutron-absorbing composite material was prepared in the same manner as in Example 1, except that the boron carbide not coated with gadolinium oxide in Example 1 was used as a reinforcing material and the content of the boron carbide was prepared according to Table 1 below.

[0039] Boron content (vol%) Comparative Example 15 Comparative Example 210 Comparative Example 315 Comparative Example 420 Comparative Example 525 Comparative Example 630

[0040] <Experimental Example 1>

[0041] FIG. 3 is a graph showing the macroscopic cross-sectional areas of Example 1 and Comparative Examples 1 to 6. The macroscopic cross-sectional area (∑) was calculated according to the following Equation 1.

[0042] [Equation 1]

[0043] ∑ = [Number of atoms per volume (N, pieces) Х Microscopic cross-sectional area (σ, cm²) 2 )] / Volume(V, cm 3 )

[0044] Example 1 and Comparative Example 3 have the same borocarbon content (15 vol%), but the macroscopic cross-sectional areas are 15.71 and 11.5 cm², respectively. -1 The macroscopic cross-sectional area of ​​Example 1 was improved by about 36.6% compared to Comparative Example 3.

[0045] Using the above macroscopic cross-sectional area value, the neutron absorption rate was calculated according to Equation 2 below.

[0046] [Equation 2]

[0047] Neutron absorption rate (%) = 1 - exp(-ΣХT)

[0048] In Equation 2 above, T is the thickness (cm) of the neutron-absorbing composite material.

[0049] The neutron absorption rates of Example 1 and Comparative Example 3 were calculated to be 90.5% and 82.1%, respectively.

[0050] <Experimental Example 2>

[0051] Figure 4 and Table 2 show the thermal conductivity measurement results of Example 1, Comparative Example 3, and Comparative Example 6. Thermal conductivity was calculated using the following Equation 3.

[0052] [Equation 3]

[0053] K=α×C p ×ρ

[0054] In Equation 3 above, K is thermal conductivity (W / mK) and α is thermal diffusivity (mm²). 2 / s), Cp is specific heat (J / g / K), and ρ is density (g / ml).

[0055] It was confirmed that Example 1 has a higher thermal conductivity than Comparative Examples 3 and 6.

[0056] Specimen density [g / ml] Thermal diffusivity [mm ２ / s]Specific heat [J / (g*K)]Thermal conductivity [W / (m*K)]30vol.% B ４ C / Al2.64647.8500.899113.8215vol.% B ４ C / Al2.67167.3300.859154.4815vol.% B ４ CARDD ２ O ３ / Al2.69967.0570.892161.44

[0057] B ４ C Thermal Conductivity: 30 W / mK

[0058] DD ２ O ３ Thermal conductivity: 2.5W / mK

[0059]

[0060] Although the present invention has been described above with reference to preferred embodiments, those skilled in the art will understand that various modifications and changes can be made to the invention without departing from the spirit and scope of the invention as set forth in the following claims.

Claims

1. Aluminum substrate; and A neutron-absorbing composite material comprising reinforcing particles dispersed within the aluminum substrate and having a boron carbide core and a gadolinium oxide shell covering the surface of the boron carbide core.

2. In Paragraph 1, A neutron-absorbing composite material having a thickness of 0.01 to 5 cm of the aluminum substrate.

3. In Paragraph 1, A neutron-absorbing composite material having an average size of 0.1 to 500 μm of the boron carbide core.

4. In Paragraph 1, A neutron-absorbing composite material having a content of 50 vol.% or less of the reinforcing particles.

5. In Paragraph 1, A neutron-absorbing composite material having a gadolinium oxide content of 2 vol.% or less.

6. In Paragraph 5, A neutron-absorbing composite material having an average thickness of 0.1 to 20 μm of the gadolinium oxide shell.

7. A step of manufacturing reinforcing powder coated with gadolinium oxide on the surface of boron carbide; A step of preparing a mixed powder by mixing the above reinforcing powder and aluminum powder; and A method for manufacturing a neutron-absorbing composite material, comprising the step of processing the above-mentioned mixed powder to manufacture a composite material in which the reinforcing powder is dispersed within an aluminum substrate.

8. In Paragraph 7, In the step of manufacturing the above storage container member, A method for manufacturing a neutron-absorbing composite material, wherein the above processing is casting or powder metallurgy.

9. Basket capable of accommodating spent nuclear fuel; A cask positioned to surround the outside of the above-mentioned container; A neutron shielding material positioned to surround the outside of the cask; and It includes an outer casing positioned to surround the outside of the above-mentioned neutron shielding material, and The above-mentioned neutron shielding material is a spent nuclear fuel storage container comprising a neutron-absorbing composite material according to claim 1.

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