Bending-resistant zirconium alloy applicable to reactor and manufacturing method thereof

Abstract

A bend-resistant zirconium alloy applicable to a reactor and a manufacturing method thereof, belonging to the field of zirconium alloys. The bend-resistant zirconium alloy applicable to a reactor includes, by weight ratio, not less than 97% of Zr, 0.61% - 1.4% of Sn, 0.6% - 1.4% of Nb, 0.05% - 0.4% of Fe, 0.1% - 0.2% of O, 0.001% - 0.005% of N, 0.001% - 0.06% of Ge; and one group of 0.001% - 0.3% of Cr and 0.01% - 0.3% of Mo and / or 0.01% - 0.09% of V and 0.01% - 0.09% of Cu; wherein, O / N ≥ 20; the structure of the alloy includes Zr(Nb,Fe,V,Cu)2 and / or Zr(Nb,Fe,Cr,Mo)2 precipitation phases. The alloy has good corrosion resistance, creep resistance and irradiation growth resistance, and has good bend resistance under high temperature, high pressure and irradiation service conditions.

Description

Technical Field

[0001] The present invention belongs to the field of zirconium alloys, and particularly relates to a bend-resistant zirconium alloy applicable to a reactor and a manufacturing method thereof. Background Art

[0002] Zirconium alloys are widely used in the cladding and structural materials of reactor fuel assemblies. At present, zirconium alloys are mainly used to manufacture the guide tubes of fuel assemblies. During service, the guide tubes transfer the axial load between the upper and lower tube seats and need to provide a certain lateral load-bearing capacity for the fuel assembly. Therefore, the zirconium alloy guide tubes need to have good bend-resistant performance under high temperature and high pressure conditions. The factors affecting the bend-resistant performance of zirconium alloys under high temperature and high pressure conditions mainly include mechanical properties, irradiation growth resistance, and corrosion resistance. At present, Zr-4 alloy, Zr-Nb alloy, E110 alloy, etc. used in civilian nuclear power plants have certain high-temperature properties, but there are deficiencies such as accelerated irradiation growth, deteriorated corrosion, and poor creep resistance under high neutron fluence rate, which affect the overall safety and economy of the reactor. Therefore, providing a bend-resistant zirconium alloy applicable to a reactor has positive significance for improving the economy of civilian nuclear power plants and enhancing the reliability of the reactor structure. Summary of the Invention

[0003] The purpose of the present invention is to provide a bend-resistant zirconium alloy applicable to a reactor to enhance the reliability of the reactor structure. The present invention also provides a manufacturing method of a bend-resistant zirconium alloy applicable to a reactor.

[0004] According to an embodiment of one aspect of the present invention, there is provided a bend-resistant zirconium alloy applicable to a reactor, which comprises, by weight ratio: not less than 97% of Zr, 0.61%-1.4% of Sn, 0.6%-1.4% of Nb, 0.05%-0.4% of Fe, 0.1%-0.2% of O, 0.001%-0.005% of N, 0.001%-0.06% of Ge; one group of 0.001%-0.3% of Cr and 0.01%-0.3% of Mo and / or 0.01%-0.09% of V and 0.01%-0.09% of Cu; wherein, O / N≥20; the structure of the bend-resistant zirconium alloy applicable to a reactor includes Zr(Nb,Fe,V,Cu)2 and / or Zr(Nb,Fe,Cr,Mo)2 precipitation phases.

[0005] Since the solubility of Sn in the α-Zr matrix is relatively large, the added Sn exists in a solid-solved form. The atomic radius of Sn (1.40 Å) is slightly smaller than that of Zr (1.45 Å). After replacing Zr atoms in the matrix, strain will be generated in the Zr lattice. The lattice strain around the solute can not only act as an obstacle to dislocation movement but also serve as an interstitial atom and vacancy trap. On the one hand, it improves the strength, and on the other hand, it can hinder the movement of point defects and dislocations during the creep process of zirconium alloys, thereby improving the creep resistance of the alloy. Adding Sn can offset the harmful effects of impurities such as N, C, and Al in sponge zirconium on the corrosion resistance. However, when the Sn content further increases, it will deteriorate the corrosion resistance of zirconium alloys. Therefore, too much Sn cannot be added. In summary, Sn can improve the mechanical properties of zirconium alloys such as yield strength, tensile strength, and creep resistance, but excessive addition will reduce the corrosion resistance of zirconium alloys. To meet the performance requirements of high burn-up fuel element guide tubes, its content needs to be strictly controlled. Considering various factors comprehensively, based on the current impurity content level in zirconium alloys and the need to improve creep resistance and corrosion resistance, the content of Sn in this invention is controlled at 0.61% - 1.4%.

[0006] A part of Nb will be solid-solved in the zirconium alloy matrix. The atomic radius of Nb (1.34 Å) is smaller than that of Zr (1.45 Å). After Nb replaces Zr, the lattice distortion generated is larger than that of Sn, so the solid-solution strengthening effect is stronger. However, the solubility of Nb is limited, and the saturated solid-solution content is 0.5%. Nb exceeding 0.5% will precipitate in the form of a second phase. By forming the Nb-containing second phase, on the one hand, it can improve the strength, and on the other hand, the Nb-containing second phase is stable after irradiation, not easily amorphized, and increases the tissue stability of the oxide film after irradiation. Nb can also make the point defects generated by irradiation recombine more effectively, so it can improve the irradiation growth resistance. However, the Nb content should not be too high. When it exceeds the saturated solid-solution concentration of 0.5%, the correct heat treatment process needs to be adopted to make the second phase distribute finely and uniformly, and then the corrosion resistance is good. In addition, too much Nb will also deteriorate the processing performance of zirconium alloys. Considering various factors comprehensively, from the need to improve the creep resistance and irradiation growth resistance of zirconium alloy guide tubes, as well as to balance the corrosion resistance and processing performance, the content of Nb in this invention is controlled at 0.6% - 1.4%.

[0007] The equilibrium solid solubility concentration of Fe in α-Zr is very low, about 300 μg / g at 730 °C, and most of the Fe precipitates in the form of the second phase. With the increase of the Fe content, more second phases will be formed in the zirconium alloy, thus being beneficial to improving the mechanical properties of the zirconium alloy. The Fe content has a significant impact on the hydrogen absorption performance of the zirconium alloy. As the Fe content in the alloy increases to about 0.5%, the hydrogen absorption fraction of the zirconium alloy decreases. Increasing the Fe content to about 0.4% can significantly improve the corrosion resistance of the alloy, but further increasing the Fe content has little effect on improving the corrosion resistance of the alloy. The Fe element can also inhibit or limit the growth of c-type dislocation loops, thus making the alloy have higher irradiation growth resistance. However, if the Fe content is too high, it is not conducive to the processing performance of the zirconium alloy. Considering various factors comprehensively, from the requirements of improving the corrosion resistance, hydrogen absorption resistance, strength and irradiation growth resistance of the zirconium alloy guide tube, as well as balancing the processing performance, the Fe content in the present invention is controlled at 0.05% - 0.4%.

[0008] Furthermore, the melting points of Cr, Ge, V, Cu, and Mo elements are > 930 °C, and the thermal neutron absorption cross sections are ≤ 5 barns. It is easy to form fine and dispersed second phases in the zirconium alloy, which hinder the movement of dislocations, thereby improving the creep resistance of the zirconium alloy.

[0009] By controlling the O content at a relatively high level of 0.1% - 0.2%, adding Sn and trace N elements at the same time, and controlling O / N ≥ 20, an O ion vacancy - N ion - Sn ion combination can be formed in the Zr alloy, so that the O ion vacancies are pinned, the vacancy migration rate is reduced, and further the nucleation of voids after irradiation is inhibited, improving the creep resistance and irradiation growth resistance of the Zr alloy. When the O content is too low, it is not conducive to the creep resistance of the Zr alloy, while when the N content is too high, it is not conducive to the corrosion resistance of the Zr alloy.

[0010] Sn, Nb, and O form a solid solution in the zirconium alloy matrix, which can hinder the movement of point defects and dislocations in the alloy after irradiation, improving the irradiation creep resistance of the Zr alloy; at the same time, when the Sn and O contents in the Zr alloy are relatively high, Nb can make the point defects generated by irradiation recombine more effectively, thereby improving the irradiation growth resistance of the Zr alloy.

[0011] By adding 0.05% - 0.4% of Fe in the Zr alloy, the growth of c-type dislocation loops can be inhibited, and the irradiation growth resistance can be improved.

[0012] Zr(Nb,Fe,V,Cu)2 and Zr(Nb,Fe,Cr,Mo)2 belong to the AB2 type Laves phase. When the Laves phase does not precipitate, it means that the thermodynamic process experienced by the zirconium alloy is unreasonable, and there may be β-Zr phase in the alloy that is not conducive to the corrosion resistance.

[0013] The Fe element can inhibit or restrict the growth of c-type dislocation loops. The Cr element has a slow diffusion rate in zirconium alloys and will stay near the original second phase, making the c-type dislocation loops in the alloy more stable. In addition, adding the Cr element can improve the corrosion resistance of zirconium alloys, including uniform corrosion resistance and weld corrosion resistance. The Mo element can, on the one hand, produce a solid solution strengthening effect, and on the other hand, can also play a strengthening role in the form of precipitation phases. Mo can increase the number of precipitation phases and reduce the size of the second-phase particles, thereby improving the anti-irradiation creep, anti-irradiation growth performance and corrosion resistance of zirconium alloys.

[0014] Because the Fe element can inhibit or restrict the growth of c-type dislocation loops, and the precipitation phases formed by Zr, Fe, and V elements have good irradiation stability and do not undergo amorphization even when irradiated to high burnup, the alloy thus has good anti-irradiation growth performance. The Cu element can increase the strength of the alloy and obtain appropriate corrosion resistance. Considering the coupling effect between alloy elements, by co-adding Nb, Fe, V, and Cu elements, Zr(Nb,Fe,V,Cu)2 is formed in the alloy. This second phase can pin the movement of grain boundaries. The above element compositions can improve the corrosion resistance of the alloy, and adding Cu in Nb-containing zirconium alloys is more effective in improving corrosion resistance than adding other alloy elements.

[0015] Furthermore, in some embodiments, by weight, the anti-bending zirconium alloy applicable to reactors contains not less than 97% Zr, 0.7%-0.9% Sn, 0.7%-0.9% Nb, 0.25%-0.35% Fe, 0.11%-0.16% O, 0.001%-0.005% N, 0.001%-0.09% V, 0.001%-0.09% Cu; and 0.001%-0.06% Ge.

[0016] The embodiments use a medium content of Sn (0.7%-0.9%) to improve corrosion resistance, a medium content of Nb (0.7%-0.9%) and O (0.11%-0.16%) to ensure the anti-creep performance of the alloy, add an appropriate amount of Fe (0.25%-0.35%) to ensure the anti-irradiation growth performance and hydrogen absorption performance of the alloy, and at the same time add a trace amount of N (0.001%-0.005%) to improve the anti-irradiation growth and anti-irradiation creep performance of the alloy. To make up for the adverse effect of the reduction in the content of Sn and Nb on strength, a trace amount of V (0.001%-0.09%) and Cu (0.001%-0.09%) are added at the same time. In addition, considering the coupling effect between alloy elements, by co-adding Nb, Fe, V, and Cu elements, Zr(Nb,Fe,V,Cu)2 is formed in the alloy. This second phase can pin the movement of grain boundaries, thereby improving the corrosion resistance of the alloy.

[0017] A trace amount of Ge can form a second phase and reduce anion vacancies, thereby improving the strength and corrosion resistance of the alloy.

[0018] Further, in some embodiments, by weight, the bend-resistant zirconium alloy applicable to a reactor contains not less than 97% of Zr, 0.61% - 1.40% of Sn, 0.6% - 1.4% of Nb, 0.1% - 0.4% of Fe, 0.11% - 0.20% of O, 0.001% - 0.005% of N, 0.001% - 0.30% of Cr, 0.001% - 0.09% of Mo; and 0.001% - 0.06% of Ge.

[0019] The embodiments adopt a medium content of Sn (0.61% - 1.0%) to improve the corrosion resistance, and a higher content of Nb (1.00% - 1.40%) and O (0.11% - 0.20%) to improve the creep resistance and irradiation growth resistance. To improve the corrosion resistance and hydrogen absorption resistance, while ensuring good workability, an appropriate amount of Fe element (0.10% - 0.4%) is added. At the same time, a trace amount of N (0.001% - 0.005%) is added to improve the irradiation growth resistance and irradiation creep resistance of the alloy. A small amount of Cr element (0.01% - 0.3%) is added to improve the corrosion resistance of the alloy. A trace amount of Mo element (0.001% - 0.09%) is added to improve the irradiation creep resistance, irradiation growth resistance and corrosion resistance of the zirconium alloy. In addition, considering the coupling effect between alloy elements, by adding Nb, Fe, Cr, and Mo elements synergistically, Zr(Nb,Fe,Cr,Mo)2 is formed in the alloy. This second phase can pin the movement of grain boundaries and inhibit grain growth, thereby improving the corrosion resistance and creep resistance of the alloy.

[0020] Further, in some embodiments, by weight, the bend-resistant zirconium alloy applicable to a reactor contains not less than 97% of Zr, 0.61% - 1.00% of Sn, 1.0% - 1.4% of Nb, 0.1% - 0.4% of Fe, 0.11% - 0.2% of O, 0.001% - 0.005% of N, 0.01% - 0.30% of Cr, 0.001% - 0.09% of Mo and 0.001% - 0.06% of Ge.

[0021] By additionally adding a trace amount of Ge, a second phase can be formed and anion vacancies can be reduced, thereby improving the strength and corrosion resistance of the alloy.

[0022] According to an embodiment of another aspect of the present invention, there is provided a manufacturing method of a bend-resistant zirconium alloy applicable to a reactor, for manufacturing the bend-resistant zirconium alloy applicable to a reactor provided in any of the foregoing embodiments. The method includes the following steps:

[0023] Step a): Provide alloy raw materials according to the composition of the bend-resistant zirconium alloy applicable to the reactor, where Zr uses nuclear-grade sponge zirconium, and the remaining alloy components use nuclear-grade metal ingots, and the non-metallic element components are present in the nuclear-grade metal ingots; use a vacuum consumable arc furnace to melt the alloy raw materials at least three times to obtain a zirconium alloy ingot;

[0024] Step b): Forge the zirconium alloy ingot at 800°C - 1100°C and keep it warm for 1h - 4h. After cooling, perform shot peening treatment and machining, and then heat it to the β-phase region and keep it warm for 5min - 10min for quenching treatment at not less than 50K / s to obtain a quenched blank;

[0025] Step c): Perform hot extrusion treatment on the quenched blank at 600 - 650°C, and then perform 3 - 4 passes of cold rolling treatment, and the deformation amount of each pass of cold rolling treatment is 40% - 80%;

[0026] Perform annealing treatment between adjacent passes of cold rolling treatment, the annealing temperature is 560°C - 600°C, and the annealing time is 1h - 4h; obtain a bend-resistant zirconium alloy finished product applicable to the reactor.

[0027] Since the atomic radii of Nb (1.34Å), Fe (1.17Å), V (1.22Å), Cu (1.17Å), Cr (1.18Å), and Mo (1.30Å) are all smaller than that of Zr atoms (1.45Å), these alloying elements will form Laves phases in the zirconium alloy after being added. Both Zr(Nb,Fe,V,Cu)2 and Zr(Nb,Fe,Cr,Mo)2 belong to Laves phases, that is, AB2-type intermetallic compounds (where A is a large atom and B is a small atom). The oxidation of this type of second phase is slower than that of the zirconium matrix. After the second phase is oxidized, volume expansion will occur, resulting in stress in the oxide film and destroying the compactness of the oxide film. By forming fine and dispersed second phases, the stress generated after the oxidation of the second phase can be distributed more evenly, reducing the degree of reduction in the protection of the oxide film. The typical temperature range for the precipitation of Laves phases in zirconium alloys is 500°C to 800°C, specifically depending on the alloy type and process. Especially after aging at 550°C to 750°C, the volume fraction of Laves phases increases with the increase in temperature. In the processing process of the present invention, both the intermediate annealing and final annealing temperatures adopt relatively high temperatures, the annealing temperature range is 560°C to 600°C, and the annealing time is 1h - 4h, which can make the Laves phases precipitate more fully. Description of the Drawings

[0028] Figure 1 It is a transmission microscope photograph of a zirconium alloy in an embodiment;

[0029] Figure 2 It is Figure 1 The electron diffraction picture of the precipitated phase in

[0030] Figure 3 is Figure 1 the energy spectrum component curve of the precipitated phase in

[0031] Figure 4 the corrosion test comparison curve of the Li-containing aqueous solution for the example and the comparative example.

[0032] The purpose of the above embodiments is to illustrate the present invention in detail so that those skilled in the art can understand the technical concept of the present invention, rather than to limit the present invention. Specific embodiments

[0033] The present invention will be further described in detail below through specific embodiments in conjunction with the accompanying drawings.

[0034] The mention of "embodiment" in this article means that the specific features, structures, or characteristics described in connection with the embodiment may be included in at least one embodiment of this article. The phrase appearing in various positions in the specification does not necessarily refer to the same embodiment, nor is it limited to mutually exclusive independent or alternative embodiments. Those skilled in the art should be able to understand that the embodiments in this article can be combined with other embodiments without structural conflicts. In the description of this article, the meaning of "a plurality" is at least two.

[0035] Due to its low thermal neutron absorption cross-section, low induced radioactivity, good comprehensive properties such as corrosion resistance and hydrogen absorption, radiation resistance, and high strength at service temperature, zirconium alloy is widely used as the cladding material and structural material of nuclear reactor fuel assemblies. Among them, the guide tube, which is one of the key structural materials of the fuel assembly, is made of zirconium alloy material. The guide tube is an important part of the fuel assembly skeleton. In the entire fuel assembly structure, it plays an important role in transmitting the axial load between the upper and lower tube seats, axially positioning the grid, balancing the force exerted by the fuel rods on the grid, and providing a certain lateral load-bearing capacity for the entire fuel assembly. The guide tube also provides an insertion channel for mobile or fixed related components, such as control rod assemblies, burnable poison assemblies, resistance plug assemblies, and neutron source assemblies, provides sufficient coolant flow for them, and provides an insertion channel and a buffer for the control rod to drop the rod. Therefore, the dimensional and structural stability of zirconium alloy under irradiation conditions is very important for the safety of the reactor. Among them, the anti-bending performance of the zirconium alloy used to manufacture the guide tube is particularly important.

[0036] The main types of materials used for guide tubes include Zr-4, M5, ZIRLO, and E635 alloys. Among them, due to the poor corrosion resistance under high burnup, Zr-4 alloy usually meets the usage requirements with a maximum burnup of the fuel assembly not exceeding 48 GWd / tU. M5 alloy has good corrosion resistance under normal PWR water chemistry conditions, but is sensitive to both high lithium concentration and dissolved oxygen, and is not suitable for use as a guide tube material under high burnup due to its poor creep resistance. The properties of ZIRLO and E635 alloys are relatively balanced, but there is still room for improvement in their corrosion resistance under high burnup. The main factors affecting the anti-bending performance of zirconium alloys under high-temperature irradiation conditions include the creep resistance (especially the irradiation creep resistance), irradiation growth resistance, mechanical properties, and corrosion resistance of zirconium alloys. Existing zirconium alloys cannot fully meet the design requirements of reactor guide tubes.

[0037] To solve the above problems, an embodiment of one aspect of the present invention provides an anti-bending zirconium alloy applicable to a reactor, which has better creep resistance, irradiation growth resistance, and corrosion resistance, and can effectively improve the reliability of the zirconium alloy guide tube.

[0038] According to an embodiment of one aspect of the present invention, an anti-bending zirconium alloy applicable to a reactor is provided. The alloy, by weight ratio, includes: not less than 97% of Zr, 0.61%-1.4% of Sn, 0.6%-1.4% of Nb, 0.05%-0.4% of Fe, 0.1%-0.2% of O, 0.001%-0.005% of N, 0.001%-0.06% of Ge; one group of 0.001%-0.3% of Cr and 0.01%-0.3% of Mo and / or 0.01%-0.09% of V and 0.01%-0.09% of Cu; wherein, O / N≥20.

[0039] In a preferred embodiment, the process of manufacturing a zirconium alloy guide tube using this alloy is as follows:

[0040] Step a): Prepare alloy raw materials according to the composition of the zirconium alloy. Among them, Zr uses nuclear-grade sponge zirconium, and the remaining alloy components use nuclear-grade metal ingots. The non-metallic element components are present in the nuclear-grade metal ingots. Use a vacuum consumable arc furnace to melt the alloy raw materials, and carry out vacuum protection during melting. The melting process is carried out at least three times to ensure the uniformity of the alloy, and a zirconium alloy ingot is obtained;

[0041] Step b): Forge the zirconium alloy ingot multiple times at 800°C - 1100°C and hold for 1h - 4h. After cooling, perform shot peening and machining drilling. Subsequently, heat it to the β-phase region, hold for 5min - 10min, and perform quenching treatment at no less than 50K / s, preferably water-cooled quenching, to obtain a quenched blank. In the quenched blank, fine Widmanstatten laths and refined second-phase particles are formed. The fine second-phase particles are beneficial to improving the corrosion resistance of the zirconium alloy.

[0042] Step c): Perform hot extrusion treatment on the quenched blank at 600 - 650°C. After extrusion, perform internal and external surface treatment to remove dirt such as oxide scale, surface defects, and grease.

[0043] Subsequently, perform 3 - 4 passes of cold rolling treatment. The wall thickness reduction per pass of cold rolling is 40% - 80%, and the area reduction rate of the last pass of rolling is 60% - 80%. A smaller deformation amount is used in the last pass of rolling of the zirconium alloy to obtain a relatively large grain size and improve the creep resistance.

[0044] Annealing treatment is carried out between adjacent passes of cold rolling treatment. The annealing temperature is 560°C - 600°C, and the annealing time is 1h - 4h to enable full recrystallization of the zirconium alloy tube. After annealing treatment, remove the oxide scale and remove dirt by pickling. The annealing treatment is carried out at a higher temperature and for a longer time to enable full recrystallization of the zirconium alloy structure and form larger grains, thereby improving the creep resistance of the zirconium alloy tube.

[0045] Finally, a finished zirconium alloy guide tube suitable for a reactor with good bending resistance is obtained.

[0046] The structure of the finished zirconium alloy guide tube is based on α-Zr. In a preferred embodiment, depending on the different alloy compositions, Zr(Nb,Fe,V,Cu)2 and / or Zr(Nb,Fe,Cr,Mo)2 precipitation phases are also distributed therein.

[0047] To verify the existence of the precipitation phase, a preliminary test sample of the present invention contains 0.8% Sn, 0.8% Nb, 0.3% Fe, 0.05% V, 0.05% Cu, 0.135% O, 0.003% N by weight, with the balance being Zr and impurities. The impurities contain 0.015% Si. The transmission electron microscope photograph of its structure is as Figure 1 shown, in which precipitation phase 1 can be seen. Electron diffraction analysis of precipitation phase 1 is as Figure 2 shown. Combining with the energy spectrum analysis as Figure 3 shown, it can be determined that precipitation phase 1 is a Zr(Nb,Fe,V,Cu)2 precipitation phase.

[0048] In different embodiments, according to the needs of the service scenario, the composition of the zirconium alloy can be adjusted adaptively.

[0049] In some embodiments, by weight, the zirconium alloy contains not less than 97% Zr, 0.7% - 0.9% Sn, 0.7% - 0.9% Nb, 0.25% - 0.35% Fe, 0.11% - 0.16% O, 0.001% - 0.005% N, 0.001% - 0.09% V, 0.001% - 0.09% Cu; and 0.001% - 0.06% Ge. Ge can refine the size of the second phase and improve the uniform distribution of the second phase, thereby improving the corrosion resistance and strength of the alloy. Ge can form the second phase and reduce anion vacancies, thereby enhancing the strength and corrosion resistance of the alloy. In a specific embodiment, the elemental composition of the zirconium alloy by weight is: 0.9% Sn, 0.9% Nb, 0.35% Fe, 0.09% V, 0.09% Cu, 0.06% Ge, 0.16% O, and 0.005% N, with the balance being Zr.

[0050] In some embodiments, by weight, the zirconium alloy contains not less than 97% Zr, 0.61% - 1.40% Sn, 0.6% - 1.4% Nb, 0.1% - 0.4% Fe, 0.11% - 0.20% O, 0.001% - 0.005% N, 0.001% - 0.30% Cr, 0.001% - 0.09% Mo; and 0.001% - 0.06% Ge.

[0051] In some embodiments, by weight, the zirconium alloy contains not less than 97% Zr, 0.61% - 1.00% Sn, 1.0% - 1.4% Nb, 0.1% - 0.4% Fe, 0.11% - 0.2% O, 0.001% - 0.005% N, 0.01% - 0.30% Cr, 0.001% - 0.09% Mo, and 0.001% - 0.06% Ge. In a specific embodiment, the elemental composition of the zirconium alloy by weight is: 0.61% Sn, 1.0% Nb, 0.1% Fe, 0.01% Cr, 0.001% Mo, 0.001% Ge, 0.11% O, and 0.001% N, with the balance being Zr.

[0052] By weight, the compositions of the Zr alloys in a set of preferred embodiments and comparative examples are shown in Table 1.

[0053]

[0054] Steady-state creep tests were carried out on the examples and comparative examples respectively. The steady-state creep rate of the Zr-4 alloy was 8×10 -6 / h, the steady-state creep rate of the example is 2×10 -6 / h. Under the same test conditions, the creep resistance of the example is improved by 74.7% compared with that of the Zr-4 alloy.

[0055] Corroded in an aqueous Li solution at a pressure of 18.6 MPa and a concentration of 70 ppm at 360 °C for 190 days, the corrosion weight gain curves of the example and the comparative example are as Figure 4 shown. It can be seen that the corrosion resistance of the example is improved by 63.1% compared with that of the Zr-4 alloy.

[0056] Room temperature tensile tests were carried out on the example and the comparative example respectively. The yield strength of the Zr-4 alloy is about 370 MPa, and the tensile strength is about 500 MPa; the yield strength of the example is about 420 MPa, and the tensile strength is about 550 MPa. It can be seen that the yield strength of the example is increased by 14.7% relative to the Zr-4 alloy, and the tensile strength is increased by 9.9%.

[0057] At the same time, due to the higher content of Nb and Fe elements and the lower content of Cr and V elements in the example, and the O / N is controlled above 20, the example has better irradiation growth resistance.

[0058] Since the bending resistance of zirconium alloys under high-temperature irradiation conditions is mainly affected by creep resistance, irradiation growth resistance, mechanical properties and corrosion resistance, the Zr alloy provided by the example has significantly better bending resistance than the comparative example Zr-4 alloy.

[0059] The purpose of the above examples is to make a further detailed description of the present invention in conjunction with the accompanying drawings, so that those skilled in the art can understand the technical concept of the present invention. Within the scope disclosed by the present invention, optimizing or equivalently replacing the technical features involved, and combining the implementation manners in different embodiments without conflict in structure and principle all fall within the protection scope of the present invention.

Claims

1. A bend-resistant zirconium alloy applicable to a reactor, characterized in that, By weight, it includes: not less than 97% of Zr, 0.61% - 1.4% of Sn, 0.6% - 1.4% of Nb, 0.05% - 0.4% of Fe, 0.1% - 0.2% of O, 0.001% - 0.005% of N, 0.001% - 0.06% of Ge; One group of 0.001% - 0.3% of Cr and 0.01% - 0.3% of Mo and / or 0.01% - 0.09% of V and 0.01% - 0.09% of Cu; The sum of the weight percentages of the above elements is 100%; Among them, O / N ≥ 20, so that an O ion vacancy - N ion - Sn ion combination is formed in the bend-resistant zirconium alloy applicable to the reactor, and the O ion vacancy is pinned; The structure of the bend-resistant zirconium alloy applicable to the reactor includes Zr(Nb,Fe,V,Cu)2 and / or Zr(Nb,Fe,Cr,Mo)2 precipitation phases.

2. The anti-bending zirconium alloy applicable to a reactor according to claim 1, characterized in that, By weight, the bend-resistant zirconium alloy applicable to the reactor contains not less than 97% of Zr, 0.7% - 0.9% of Sn, 0.7% - 0.9% of Nb, 0.25% - 0.35% of Fe, 0.11% - 0.16% of O, 0.001% - 0.005% of N, 0.01% - 0.09% of V, 0.01% - 0.09% of Cu; And 0.001% - 0.06% of Ge.

3. The anti-bending zirconium alloy applicable to a reactor according to claim 1, characterized in that, By weight, the bend-resistant zirconium alloy applicable to the reactor contains not less than 97% of Zr, 0.61% - 1.40% of Sn, 0.6% - 1.4% of Nb, 0.1% - 0.4% of Fe, 0.11% - 0.20% of O, 0.001% - 0.005% of N, 0.001% - 0.30% of Cr, 0.01% - 0.09% of Mo; And 0.001% - 0.06% Ge.

4. The anti-bending zirconium alloy applicable to a reactor according to claim 3, characterized in that, By weight, the bend-resistant zirconium alloy applicable to the reactor contains not less than 97% of Zr, 0.61% - 1.00% of Sn, 1.0% - 1.4% of Nb, 0.1% - 0.4% of Fe, 0.11% - 0.2% of O, 0.001% - 0.005% of N, 0.01% - 0.30% of Cr, 0.01% - 0.09% of Mo and 0.001% - 0.06% of Ge.

5. A manufacturing method of a bend-resistant zirconium alloy applicable to a reactor, characterized in that, For manufacturing the bend-resistant zirconium alloy applicable to the reactor according to any one of claims 1 to 4, and includes the following steps: Step a): Provide alloy raw materials according to the composition of the bend-resistant zirconium alloy applicable to the reactor, where Zr is nuclear-grade sponge zirconium, and the remaining alloy components are nuclear-grade metal ingots, and the non-metal element components are present in the nuclear-grade metal ingots; Use a vacuum consumable arc furnace to melt the alloy raw materials at least three times to obtain a zirconium alloy ingot; Step b): Forge the zirconium alloy ingot at 800°C - 1100°C and hold for 1 h - 4 h, perform shot peening and machining after cooling, and then heat to the β-phase region and hold for 5 min - 10 min for quenching treatment at not less than 50 K / s to obtain a quenched blank; Step c): Perform hot extrusion treatment on the quenched blank at 600 - 650°C, and then perform 3 - 4 passes of cold rolling treatment, with the deformation amount of each cold rolling treatment being 40% - 80%; Annealing treatment is carried out between adjacent passes of cold rolling treatment, the annealing temperature is 560°C - 600°C, and the annealing time is 1 h - 4 h; Obtain a finished anti-bending zirconium alloy suitable for reactors.

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