WO2019052315A1 - Fuel cladding and fuel assembly - Google Patents

Abstract

A fuel cladding and a fuel assembly, the fuel cladding comprising a zirconium alloy substrate, an intermediate layer having a non-stoichiometric ratio and a gradient characteristic disposed on the zirconium alloy substrate, and an environmental shielding layer disposed on the intermediate layer; the intermediate layer and The environmental barrier layer forms a gradient complex coating having a non-stoichiometric ratio on the zirconium alloy substrate. It overcomes the problems of interfacial stress, interfacial diffusion and high temperature steam oxidation in the single coating of traditional zirconium alloy cladding. It is suitable for accidental fault-tolerant nuclear fuel cladding and greatly improves the nuclear fuel reactor structure under severe accident conditions. Anti-accident capability and safety threshold with functional integrity.

Description

Fuel cladding and fuel assembly Technical field

The invention relates to the technical field of nuclear reactors, and in particular to a fuel cladding and a fuel assembly.

Background technique

After some nuclear accidents, nuclear power safety has once again become the focus of international public attention. How to further improve the safety of nuclear power, especially to improve the safety threshold of nuclear reactors against ultra-designed nuclear accidents, has become an important issue for the sustainable development of nuclear energy. Accident-tolerant nuclear fuel (Accident Tolerant Fuels (ATF), a new nuclear safety technology concept, was born in this context and gradually became one of the most important research topics in the world's nuclear power industry. Its purpose is to the existing zirconium alloy/uranium dioxide fuel system. Improvements, upgrades, and even full replacements to reduce the heat and hydrogen generation of the cladding and high-temperature steam, improve the structural integrity and functionality of the cladding at 1200 °C, and enhance the containment of fission gases. Ability, etc. The application history of zirconium alloy cladding in nuclear reactors has been more than 50 years. As a material approved by nuclear power plants, improvement on the basis of this is the most practical and feasible technical route at this stage.

At present, there are two main ways to improve the surface properties of zirconium alloys: (1) coating technology, coating a layer of heterogeneous film on the surface of zirconium alloy by electroplating, electroless plating, thermal spraying, vapor deposition, etc., due to the zirconium alloy cladding Long-term use in extreme harsh environments of high temperature, high pressure, erosion, irradiation, erosion, coatings inevitably have interface bonding, thermal expansion matching and other issues; (2) surface modification technology, through surface heat treatment, chemical heat treatment, laser surface Treatment, ion implantation, etc. to change the surface morphology, chemical composition, phase composition, microstructure, defect state or stress state of the zirconium alloy surface, thereby improving its surface properties.

From the public reports, there are many studies on the alloying of niobium (Nb) in surface alloying. Lee et al. used laser alloying on the surface of Zr-4 alloy to solidify Nb into the Zr lattice, although fine crystals were formed. The structure improves the surface hardness and the corrosion resistance of the chloride solution, but in the water vapor of 400 ° C, the corrosion resistance decreases due to the formation and hydrogenation of β-Zr. In addition, the surface alloying elements and the Zr matrix are easily interdiffused, especially under high temperature conditions, and the final surface modification fails due to the development of time. In surface ceramization, the most research is to form an oxide film on the surface of zirconium alloy. For example, Russia adds anodized film on the water side. Westinghouse inductively oxidizes in air. General Electric Company carries out autoclave pre-film on the surface of the cladding. Wait. Although the oxide film (ZrO ₂ ) can improve the corrosion resistance of the zirconium alloy, ZrO ₂ is a thermal barrier material (thermal conductivity is only 1.8-3.0 W/mK), which will seriously hinder the heat between the core and the coolant. Exchange efficiency. In addition, in the high temperature conditions of the accident, ZrO ₂ has a phase change crack, and the oxidation penetrates into the zirconium matrix along the crack, which invalidates the surface modification.

In 1987, Siemens applied for a patent for fuel cladding surface coatings, including TiC, TiN, ZrN, CrC, TiAlVN, TaN, ZrC and WC. The patent mainly considers wear resistance and normal work. The hydrothermal corrosion performance under the condition does not take into account the high-temperature steam oxidation performance under the conditions of water loss accidents. The high-temperature steam oxidation performance of these coating systems is poor.

In summary, the research on surface modification of zirconium alloy is still limited to hydrothermal corrosion, hydrogen absorption, wear resistance, etc., and does not comprehensively consider accident conditions (such as 1200 °C high temperature caused by water loss accident), heat exchange. Efficiency, radiation resistance, and compatibility of the coating with the zirconium alloy matrix (eg, lattice matching, thermal conductivity matching, thermal expansion matching, etc.).

technical problem

The technical problem to be solved by the present invention is to provide a fuel cladding and an fuel assembly having the fuel cladding that improve the anti-accident capability and the safety threshold.

Technical solution

The technical solution adopted by the present invention to solve the technical problem thereof is to provide a fuel cladding comprising a zirconium alloy substrate, an intermediate layer having non-stoichiometric ratios and gradient characteristics disposed on the zirconium alloy substrate, and An environmental barrier layer on the intermediate layer; the intermediate layer and the environmental barrier layer form a gradient complex coating having a non-stoichiometric ratio on the zirconium alloy substrate.

In the fuel cladding of the present invention, the gradient multiphase coating has a density of from 90% to 100% and a porosity of from 10% to 0%.

In the fuel cladding of the present invention, the intermediate layer is one or more of a ZrC _1-x coating, a ZrN _1-x coating, a TiC _1-x coating, and a TiN _1-x coating, wherein x is 0-0.5.

In the fuel cladding of the present invention, the intermediate layer has a thickness of from 0.1 μm to 10 μm.

In the fuel cladding of the present invention, the environmental shielding layer is one or more of a SiC coating, a MAX phase coating, and a CrN coating.

In the fuel cladding of the present invention, the MAX phase coating is Ti ₃ SiC, Ti3AlC ₂ , Ti ₂ AlC, Cr ₂ AlC, Ti ₂ AlN, Zr ₃ SiC, Zr ₃ AlC ₂ , Zr ₂ AlN and Cr ₂ AlN. One or more.

In the fuel cladding of the present invention, the environmental shielding layer has a thickness of from 0.1 μm to 100 μm.

In the fuel cladding of the present invention, the portion of the coating where the intermediate layer and the environmental shield are joined together form a transition layer.

In the fuel cladding of the present invention, the intermediate layer and the environmental barrier layer are respectively formed on the surface of the zirconium alloy substrate by physical vapor deposition.

The invention also provides a fuel assembly comprising the fuel cladding of any of the above.

Beneficial effect

The invention has the beneficial effects of overcoming the problems of interfacial stress, interfacial diffusion and high temperature steam oxidation in the single coating of the traditional zirconium alloy cladding. By co-design, the fuel cladding of the invention has a non-stoichiometric gradient. The multi-phase coating is suitable for accident-tolerant nuclear fuel cladding applications, greatly improving the anti-accident capability and safety threshold of nuclear reactors to maintain the structural and functional integrity of nuclear fuel assemblies under severe accident conditions.

DRAWINGS

The present invention will be further described below in conjunction with the accompanying drawings and embodiments, in which:

1 is a cross-sectional structural view showing a fuel cladding according to an embodiment of the present invention;

2 is a high resolution transmission electron micrograph of a grain boundary of a gradient composite coating in the fuel cladding embodiment 1 of the present invention;

Figure 3 is a graph showing the lattice volume swelling of the intermediate layer under different irradiation conditions in Example 1 of the present invention;

Figure 4 is a cross-sectional scanning electron micrograph of a gradient multiphase coating of the fuel cladding embodiment 2 of the present invention;

Fig. 5 is a front view and front view of the intermediate layer in the second embodiment of the present invention.

Embodiments of the invention

For a better understanding of the technical features, objects and effects of the present invention, the embodiments of the present invention are described in detail with reference to the accompanying drawings.

As shown in FIG. 1, the fuel cladding of the present invention comprises a zirconium alloy substrate 10, an intermediate layer 20 having non-stoichiometric ratios and gradient characteristics disposed on the zirconium alloy substrate 10, and an environmental shield disposed on the intermediate layer 20. Layer 30; the zirconium alloy substrate 10 is the body of the fuel cladding, and the intermediate layer 20 and the environmental shielding layer 30 form a gradient complex coating having a non-stoichiometric ratio on the zirconium alloy substrate 10.

The zirconium alloy substrate 10 is generally of a tubular structure, and FIG. 1 only shows a laminated structure of a fuel cladding, and the zirconium alloy substrate 10 is also only a partial structure. The gradient composite coating is disposed on the surface (outer surface) of the zirconium alloy substrate 10. The intermediate layer 20 is located between the zirconium alloy substrate 10 and the environmental shielding layer 30, and has non-stoichiometric characteristics and gradient characteristics. First, the difference between the large thermal expansion coefficient of the environmental shielding layer 30 and the zirconium alloy substrate 10 is alleviated. It is to prevent the interface diffusion reaction between the environmental shielding layer 30 and the zirconium alloy substrate 10 at high temperature, and the third is that the lattice vacancy of the non-stoichiometric transition layer can play the role of self-healing of the radiation damage defect, avoiding the irradiation environment. Stress cracking caused by interface damage between the undercoat layer and the zirconium alloy substrate 10.

The intermediate layer 20 may be one or more of a ZrC _1-x coating, a ZrN _1-x coating, a TiC _1-x coating, and a TiN _1-x coating, where x is 0-0.5.

The gradient feature of the intermediate layer 20 is mainly represented by a composition gradient; for the intermediate layer 20 having a plurality of component coatings, each coating layer may be subjected to a gradient distribution of components or a gradient of concentration gradients. For example, for the intermediate layer 20 of the ZrC _1-x coating, where C has a plurality of concentrations by the difference in value of X, the ZrC _1-x coating may vary from less to more or less depending on the C concentration. distributed.

Alternatively, the intermediate layer 20 has a thickness of from 0.1 μm to 10 μm.

The environmental shielding layer 30 is located on the outer side, and the environmental shielding layer 30 has excellent high-temperature oxidation resistance and wear resistance, and protects the high-temperature oxidation of the zirconium alloy fuel cladding under accident conditions and resists the fretting wear of the grid.

The environmental shielding layer 30 may be one or more of a SiC coating, a MAX phase coating, and a CrN coating, which has high thermal conductivity, high strength, high radiation tolerance, corrosion resistance, and accident resistance. High temperature steam oxidation, wear resistance and so on. Wherein the MAX phase coating layer may be one or more of Ti ₃ SiC, Ti ₃ AlC ₂ , Ti ₂ AlC, Cr ₂ AlC, Ti ₂ AlN, Zr ₃ SiC, Zr ₃ AlC ₂ , Zr ₂ AlN and Cr ₂ AlN. Kind.

Alternatively, the thickness of the environmental shielding layer 30 is from 0.1 μm to 100 μm.

Further, the coating portions where the intermediate layer 20 and the environmental shielding layer 30 meet are combined to form the transition layer 40. The composition of the transition layer 40 is a combination of components of the intermediate layer 20 and the environmental shielding layer 30. For example, when the intermediate layer 20 is a ZrC _1-x coating and the environmental shielding layer 30 is a SiC coating, the transition layer 40 formed by the combination of the two is a SiC-ZrC _1-x layer.

In the fuel cladding, the intermediate layer 20 and the environmental shielding layer 30 are respectively formed on the surface of the zirconium alloy substrate by physical vapor deposition to form a gradient composite coating. The gradient composite coating has a density of from 90% to 100% and a porosity of from 10% to 0%.

The fuel assembly of the present invention includes the fuel cladding described above.

The invention is further illustrated by the following examples.

Example 1

By physical vapor deposition, a ZrC _0.7 intermediate layer of 0.5 μm thickness is first deposited on the surface of the zirconium alloy substrate, and a SiC environmental shielding layer is deposited on the ZrC _0.7 intermediate layer. The thickness of the SiC environment shielding layer is 2 μm. The ZrC _0.7 intermediate layer and the SiC environmental shielding layer have a density of >99%, a porosity of <1%, and a bonding strength of the coating to the zirconium alloy matrix of >70 MPa. A high-resolution projection electron micrograph of the ZrC _0.7 / SiC gradient composite coating grain boundary is shown in Fig. 2.

In terms of high temperature oxidation resistance, the zirconium alloy substrate having the gradient composite coating has an oxidative weight gain of only 0.2 mg/cm ² by steam oxidation at 1200 ° C for 1 hour, while the uncoated zirconium alloy matrix is the same. The oxidative weight gain under the condition is 37 mg/cm ² , which indicates that the gradient composite coating effectively reduces the high temperature steam oxidation weight gain of the zirconium alloy nuclear fuel cladding by two orders of magnitude.

In the interface diffusion reaction between the coating and the zirconium alloy matrix, the ZrC _1-x intermediate layer design, the zirconium alloy substrate and the coating are kept at a high temperature of 1200 ° C for 30 minutes without obvious diffusion reaction at the interface, compared with the conventional uncoated zirconium. The alloy cladding increases the 400 °C withstand temperature. As shown in Figure 3, in the anti-radiation damage, the introduction and regulation of non-stoichiometric ZrC _0.7 intermediate layer carbon vacancies at 7 MeV high energy Xe ²⁶⁺ ion 700 ° C high temperature irradiation 2.5 × 10 ¹⁵ /cm ² The lattice expansion is less than 1%, and it can be seen that the non-stoichiometric ZrC _0.7 achieves self-healing of the radiation damage defect.

Example 2

By physical vapor deposition, a TiN _0.7 intermediate layer of 1 μm thickness was first deposited on the surface of the zirconium alloy substrate, and a Cr ₂ AlN environmental shielding layer was deposited on the TiN _0.7 intermediate layer. The thickness of the Cr ₂ AlN environmental shielding layer was 1 μm. The TiN _0.7 intermediate layer and the Cr ₂ AlN environmental shielding layer have a density of >99%, a porosity of <1%, and a bonding strength of the coating to the zirconium alloy matrix of >60 MPa. A cross-sectional scanning electron micrograph of a TiN _0.7 /Cr ₂ AlN gradient composite coating is shown in FIG.

In terms of high temperature oxidation resistance, the zirconium alloy substrate having the gradient composite coating has an oxidative weight gain of only 0.6 mg/cm ² by steam oxidation at 1200 ° C for 1 hour, while the uncoated zirconium alloy cladding is The oxidative weight gain under the same conditions was 37 mg/cm ² , indicating that the gradient composite coating effectively reduced the high temperature steam oxidation weight gain of the zirconium alloy nuclear fuel cladding by two orders of magnitude.

In the interface diffusion reaction between the coating and the zirconium alloy matrix, through the TiN _1-x intermediate layer design, the zirconium alloy substrate and the coating are incubated at 1200 ° C for 30 minutes without significant diffusion reaction at the interface, compared with the conventional uncoated zirconium. The alloy cladding increases the 400 °C withstand temperature, while the Cr ₂ AlN without the intermediate layer undergoes a significant diffusion reaction with the zirconium alloy. In terms of radiation damage resistance, the lattice constant did not change significantly by introducing and regulating the nitrogen vacancies in the non-stoichiometric TiN _0.7 intermediate layer at 300 °C Ar ion 800 °C high temperature irradiation 3 × 10 ¹⁷ /cm ² . It can be seen that non-stoichiometric TiN _0.7 achieves self-healing of radiation damage defects.

The non-stoichiometric TiN _0.7 intermediate layer and the stoichiometric TiN intermediate layer of the present invention are compared before and after irradiation, and the lattice constant changes of the two are shown in the figure. In FIG. 5, TiN is a stoichiometric ratio unirradiated sample, i -TiN is a stoichiometric post-irradiation sample, TiN _{0.7 is a} non-stoichiometric unirradiated sample, and i-TiN _{0.7 is a} non-stoichiometric post-irradiation sample. As can be seen from the figure, the lattice constant of the non-stoichiometric TiN _0.7 intermediate layer did not change significantly.

The above is only the embodiment of the present invention, and is not intended to limit the scope of the invention, and the equivalent structure or equivalent process transformation of the present invention and the contents of the drawings may be directly or indirectly applied to other related technologies. The fields are all included in the scope of patent protection of the present invention.

Claims (10)

A fuel cladding characterized by comprising a zirconium alloy substrate, an intermediate layer having non-stoichiometric ratios and gradient characteristics disposed on the zirconium alloy substrate, and an environmental shielding layer disposed on the intermediate layer; The intermediate layer and the environmental barrier layer form a gradient multiphase coating having a non-stoichiometric ratio on the zirconium alloy substrate. The fuel cladding according to claim 1 wherein said gradient multiphase coating has a density of from 90% to 100% and a porosity of from 10% to 0%. The fuel cladding according to claim 1, wherein said intermediate layer is one of a ZrC _1-x coating, a ZrN _1-x coating, a TiC _1-x coating, and a TiN _1-x coating. One or more, where x is 0-0.5. The fuel cladding according to claim 1, wherein the intermediate layer has a thickness of from 0.1 μm to 10 μm. The fuel cladding of claim 1 wherein said environmental barrier layer is one or more of a SiC coating, a MAX phase coating, and a CrN coating. The fuel cladding according to claim 5, wherein the MAX phase coating is Ti ₃ SiC, Ti3AlC ₂ , Ti ₂ AlC, Cr ₂ AlC, Ti ₂ AlN, Zr ₃ SiC, Zr ₃ AlC ₂ , Zr ₂ One or more of AlN and Cr ₂ AlN. The fuel cladding according to claim 1, wherein the environmental shielding layer has a thickness of from 0.1 μm to 100 μm. A fuel cladding according to claim 1 wherein the portion of the coating where the intermediate layer and the environmental shield are joined together form a transition layer. The fuel cladding according to any one of claims 1 to 8, wherein the intermediate layer and the environmental shielding layer are respectively formed on the surface of the zirconium alloy substrate by physical vapor deposition. A fuel assembly characterized by comprising the fuel cladding of any of claims 1-9.