Bending-resistant zirconium alloy suitable for reactor and manufacturing method of bending-resistant zirconium alloy

Abstract

The invention discloses a bending-resistant zirconium alloy suitable for a reactor and a manufacturing method thereof, and belongs to the field of zirconium alloys. The invention relates to a bending-resistant zirconium alloy suitable for a reactor. The bending-resistant zirconium alloy comprises the following components in percentage by weight: 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 and 0.001%-0.06% of Ge. And one of 0.001% to 0.3% of Cr, 0.01% to 0.3% of Mo, and / or 0.01% to 0.09% of V, and 0.01% to 0.09% of Cu; wherein O / N is greater than or equal to 20; the structure of the alloy comprises Zr (Nb, Fe, V, Cu) 2 and / or Zr (Nb, Fe, Cr, Mo) 2 precipitated phases. The alloy has good corrosion resistance, creep resistance and irradiation growth resistance, and has good bending resistance under high-temperature and high-pressure irradiation service conditions.

Description

Technical Field

[0001] The invention belongs to the field of zirconium alloys, and in particular relates to a bending-resistant zirconium alloy suitable for a reactor and a manufacturing method thereof. Background Art

[0002] Zirconium alloys are widely used in reactor fuel assembly cladding and structural materials. Fuel assembly guide tubes are currently mainly made of zirconium alloys. During service, the guide tubes transmit 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 bending resistance under high temperature and high pressure conditions. The factors affecting the bending resistance of zirconium alloys under high temperature and high pressure conditions mainly include mechanical properties, radiation growth resistance and corrosion resistance. At present, the Zr-4 alloy, Zr-Nb alloy, E110 alloy, etc. used in civilian nuclear power plants have certain high temperature properties, but there are shortcomings such as accelerated radiation growth, corrosion deterioration, and poor creep resistance under high neutron injection rates, which affect the overall safety and economy of the reactor. Therefore, providing a bending-resistant zirconium alloy suitable for reactors is of positive significance for improving the economy of civilian nuclear power plants and improving the reliability of reactor structures. Summary of the invention

[0003] The purpose of the present invention is to provide a bending-resistant zirconium alloy suitable for a reactor to improve the reliability of the reactor structure. The present invention also provides a method for manufacturing the bending-resistant zirconium alloy suitable for a reactor.

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

[0005] Since Sn has a large solid solubility in the α-Zr matrix, the added Sn exists in the form of solid solution. 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 serve as an obstacle to dislocation movement, but also act as interstitial atoms and vacancy traps. On the one hand, it improves the strength, and on the other hand, it can hinder the movement of point defects and dislocations in 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 corrosion resistance, but further increasing the Sn content will deteriorate the corrosion resistance of zirconium alloys, so it cannot be added too much. 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. In order to meet the performance requirements of the guide tube of high-burnup fuel elements, its content needs to be strictly controlled. Taking various factors into consideration, based on the current level of impurities in zirconium alloys and the need to improve creep resistance and corrosion resistance, the Sn content in the present invention is controlled within a range of 0.61%-1.4%.

[0006] Part of Nb will be dissolved 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 is larger than that of Sn, so the solid solution strengthening effect is stronger. However, the solid 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 a second phase containing Nb, on the one hand, the strength can be improved. On the other hand, the second phase containing Nb is stable after irradiation and is not easy to produce amorphization, which increases the organizational stability of the oxide film after irradiation. Nb can also make the point defects generated by irradiation more effectively compound, thereby improving the radiation 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 must be adopted to make the second phase small and evenly distributed, so that the corrosion resistance is good. In addition, too much Nb will also deteriorate the processing performance of the zirconium alloy. Taking all factors into consideration, in order to improve the creep resistance and radiation growth resistance of the zirconium alloy guide tube and balance the corrosion resistance and processing performance, the Nb content in the present invention is controlled within a range of 0.6%-1.4%.

[0007] The equilibrium solid solution concentration of Fe in α-Zr is very low, about 300 μg / g at 730°C, and most of Fe precipitates in the form of a second phase. As the Fe content increases, more second phases will be formed in the zirconium alloy, which is beneficial to improving the mechanical properties of the zirconium alloy. The Fe content has a significant effect 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 rings, thereby making the alloy have higher resistance to radiation growth. However, too high a Fe content is not conducive to the processing performance of the zirconium alloy. Taking all factors into consideration, from the requirements of improving the corrosion resistance, hydrogen absorption resistance, strength and radiation 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 greater than 930°C, and their thermal neutron absorption cross sections are less than or equal to 5 targets. They are prone to form a fine dispersed second phase in the zirconium alloy, hindering dislocation movement, thereby improving the creep resistance of the zirconium alloy.

[0009] By controlling the O content at a 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 vacancy is pinned, the vacancy mobility is reduced, and the nucleation of the void after irradiation is inhibited, and the creep resistance and radiation growth resistance of the Zr alloy are improved. 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, and improve the radiation creep resistance of the Zr alloy; at the same time, when the Sn and O contents in the Zr alloy are high, Nb can make the point defects generated by irradiation more effectively recombine, thereby improving the radiation growth resistance of the Zr alloy.

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

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

[0013] Fe can inhibit or limit the growth of c-type dislocation loops, while Cr diffuses slowly in zirconium alloys and stays near the original second phase, making the c-type dislocation loops in the alloy more stable. In addition, the addition of Cr can improve the corrosion resistance of zirconium alloys, including resistance to uniform corrosion and welding corrosion. Mo can produce a solid solution strengthening effect on the one hand, and on the other hand, it can also play a reinforcing role in the form of precipitation phases. Mo can increase the number of precipitation phases and reduce the size of second phase particles, thereby improving the radiation creep resistance, radiation growth resistance and corrosion resistance of zirconium alloys.

[0014] Because the Fe element can inhibit or limit the growth of c-type dislocation rings, and the precipitated phase formed by the Zr, Fe, and V elements has good radiation stability, and does not undergo amorphization when irradiated to high burnup, the alloy has good resistance to radiation growth. The Cu element can improve the strength of the alloy and obtain suitable corrosion resistance. Considering the coupling effect between alloying elements, Zr(Nb,Fe,V,Cu)2 is formed in the alloy by synergistically adding Nb, Fe, V, and Cu elements. This second phase can pin the movement of grain boundaries. The above elemental composition can improve the corrosion resistance of the alloy, and adding Cu to Nb-containing zirconium alloys is more effective in improving corrosion resistance than adding other alloying elements.

[0015] Furthermore, in some embodiments, the bending-resistant zirconium alloy suitable for the reactor contains, by weight, 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 embodiment adopts a medium content of Sn (0.7%-0.9%) to improve the corrosion resistance, a medium content of Nb (0.7%-0.9%) and O (0.11%-0.16%) to ensure the creep resistance of the alloy, and an appropriate amount of Fe (0.25%-0.35%) is added to ensure the alloy's radiation growth resistance and hydrogen absorption performance, while adding a trace amount of N (0.001%-0.005%) to improve the alloy's radiation growth resistance and radiation creep resistance. In order to compensate for the adverse effects of reduced Sn and Nb content 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 alloying elements, Zr (Nb, Fe, V, Cu) 2 is formed in the alloy by synergistically adding Nb, Fe, V, and Cu elements. This second phase can pin the movement of grain boundaries, thereby improving the corrosion resistance of the alloy.

[0017] Trace amounts of Ge can form a second phase, reduce anion vacancies, and thus improve the strength and corrosion resistance of the alloy.

[0018] Furthermore, in some embodiments, the bending-resistant zirconium alloy suitable for the reactor contains, by weight, 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.

[0019] The embodiment adopts a medium content of Sn (0.61%-1.0%) to improve the corrosion resistance, and a relatively high content of Nb (1.00%-1.40%) and O (0.11%-0.20%) to improve the creep resistance and radiation growth resistance. In order to improve the corrosion resistance and hydrogen absorption resistance while ensuring good processing performance, an appropriate amount of Fe element (0.10%-0.4%) is added, and a trace amount of N (0.001%-0.005%) is added to improve the alloy's resistance to radiation growth and radiation creep. 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 radiation creep resistance, radiation growth resistance and corrosion resistance of the zirconium alloy. In addition, considering the coupling effect between alloy elements, Zr(Nb,Fe,Cr,Mo)2 is formed in the alloy by synergistically adding Nb, Fe, Cr and Mo elements. 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] Furthermore, in some embodiments, the bending-resistant zirconium alloy suitable for the reactor contains, by weight, 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.

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

[0022] According to another embodiment of the present invention, a method for manufacturing a bending-resistant zirconium alloy suitable for a reactor is provided, which is used to manufacture the bending-resistant zirconium alloy suitable for a reactor provided in any of the above embodiments. The method comprises the following steps:

[0023] Step a): providing alloy raw materials according to the composition of the bending-resistant zirconium alloy suitable for the reactor, wherein Zr is nuclear-grade sponge zirconium, and the other alloy components are nuclear-grade metal ingots, and non-metallic elements are present in the nuclear-grade metal ingots; using a vacuum consumable arc furnace to melt the alloy raw materials at least three times to obtain a zirconium alloy ingot;

[0024] Step b): forging the zirconium alloy ingot at 800°C-1100°C and keeping it warm for 1h-4h, performing shot blasting and machining after cooling, and then heating it to the β phase region and keeping it warm for 5min-10min to perform quenching treatment at not less than 50K / s to obtain a quenched billet;

[0025] Step c): hot extruding the quenched blank at 600-650° C., followed by 3-4 passes of cold rolling, with the deformation of each pass of cold rolling being 40%-80%;

[0026] Annealing treatment is performed between adjacent cold rolling treatments, the annealing temperature is 560°C-600°C, and the annealing time is 1h-4h; and a bending-resistant zirconium alloy product suitable for a reactor is obtained.

[0027] Since the atomic radius 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 zirconium alloys after addition. Zr(Nb,Fe,V,Cu)2 and Zr(Nb,Fe,Cr,Mo)2 are both Laves phases, i.e., AB2 type intermetallic compounds (where A is a large atom and B is a small atom). This type of second phase oxidizes slower than the zirconium matrix. After the second phase is oxidized, it will produce volume expansion, resulting in stress in the oxide film and destroying the compactness of the oxide film. By forming a fine and dispersed second phase, the stress generated after the second phase is oxidized can be distributed more evenly, reducing the degree of reduction in the protectiveness of the oxide film. The typical temperature range for the precipitation of Laves phase in zirconium alloy is 500°C~800°C, which depends on the alloy type and process. In particular, after aging at 550°C~750°C, the volume fraction of Laves phase increases with increasing temperature. The intermediate annealing and final annealing temperatures in the processing of the present invention are both relatively high. The annealing temperature range is 560°C~600°C, and the annealing time is 1h-4h, which can make the Laves phase precipitate more fully. BRIEF DESCRIPTION OF THE DRAWINGS

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

[0029] Figure 2 for Figure 1 Electron diffraction image of the precipitated phase;

[0030] Figure 3 for Figure 1 Energy spectrum composition curve of the precipitated phase;

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

[0032] The purpose of the above embodiments is to explain the present invention in detail so that those skilled in the art can understand the technical concept of the present invention, but it is not intended to limit the present invention. DETAILED DESCRIPTION

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

[0034] Reference to "embodiment" herein means that a particular feature, structure or characteristic described in conjunction with the embodiment may be included in at least one embodiment of the present invention. The phrase appearing in various locations in the specification does not necessarily refer to the same embodiment, nor is it limited to mutually exclusive independent or alternative embodiments. It should be understood by those skilled in the art that the embodiments herein may be combined with other embodiments without structural conflicts. In the description herein, "multiple" means at least two.

[0035] Zirconium alloys are widely used as cladding materials and structural materials for nuclear reactor fuel assemblies due to their low thermal neutron absorption cross section, low induced radioactivity, corrosion resistance, hydrogen absorption, radiation resistance and high strength at service temperature. Among them, the guide tube, one of the key structural materials of the fuel assembly, is made of zirconium alloy. 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, axial positioning of the grid, balancing the force exerted by the fuel rod 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, providing them with sufficient coolant flow, and providing an insertion channel and a buffering effect for the control rods. Therefore, the dimensional and structural stability of zirconium alloys under irradiation conditions is very important for the safety of the reactor. Among them, the bending resistance of the zirconium alloy used to manufacture the guide tube is particularly important.

[0036] The types of materials used for guide tubes mainly include Zr-4, M5, ZIRLO and E635 alloys. Among them, Zr-4 alloy has poor corrosion resistance under high fuel consumption, and usually meets the use requirements of fuel assemblies with a maximum fuel consumption of no more than 48GWd / tU. M5 alloy has good corrosion resistance under normal PWR water chemistry conditions, but is sensitive to high lithium concentrations and dissolved oxygen, and is not suitable for use as a guide tube material under high fuel consumption due to its poor creep resistance. The various properties of ZIRLO and E635 alloys are relatively balanced, but their corrosion resistance under high fuel consumption still has room for improvement. The main factors affecting the bending resistance of zirconium alloys under high-temperature irradiation conditions include the creep resistance of zirconium alloys (especially radiation creep resistance), radiation growth resistance, mechanical properties and corrosion resistance. Existing zirconium alloys cannot fully meet the design requirements of reactor guide tubes.

[0037] In order to solve the above problems, an embodiment of one aspect of the present invention provides a bending-resistant zirconium alloy suitable for a reactor, which has better creep resistance, radiation 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, there is provided a bending-resistant zirconium alloy suitable for a reactor, the alloy comprising, by weight: not less than 97% Zr, 0.61%-1.4% Sn, 0.6%-1.4% Nb, 0.05%-0.4% Fe, 0.1%-0.2% O, 0.001%-0.005% N, 0.001%-0.06% Ge; 0.001%-0.3% Cr and a group selected from 0.01%-0.3% Mo and / or 0.01%-0.09% V and 0.01%-0.09% Cu; wherein O / N≥20.

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

[0040] Step a): preparing alloy raw materials according to the composition of zirconium alloy, wherein Zr adopts nuclear grade sponge zirconium, the other alloy components adopt nuclear grade metal ingots, and non-metallic elements exist in the nuclear grade metal ingots. The alloy raw materials are melted in a vacuum consumable arc furnace, vacuum protection is performed during the melting, and the melting process is performed at least three times to ensure the uniformity of the alloy, and a zirconium alloy ingot is obtained;

[0041] Step b): The zirconium alloy ingot is forged multiple times at 800°C-1100°C and kept warm for 1h-4h, and after cooling, it is shot blasted and machined and drilled, and then heated to the β phase region and kept warm for 5min-10min for quenching at not less than 50K / s, preferably water-cooled quenching, to obtain a quenched blank. Small-sized Widmanstattenite laths and refined second phase particles are formed in the quenched blank. The fine second phase particles are conducive to improving the corrosion resistance of the zirconium alloy.

[0042] Step c): hot extruding the quenched blank at 600-650° C., and treating the inner and outer surfaces after extrusion to remove scale, surface defects, grease and other contaminants.

[0043] Subsequently, 3-4 cold rolling passes are carried out, with each pass reducing the wall thickness by 40%-80%, and the last pass reducing the area by 60%-80%. The last pass of zirconium alloy rolling uses a smaller deformation to obtain a relatively larger grain size to improve creep resistance.

[0044] Annealing treatment is performed between adjacent cold rolling treatments, with an annealing temperature of 560℃-600℃ and an annealing time of 1h-4h to allow the zirconium alloy tube to fully recrystallize. After annealing treatment, the oxide scale is removed and the dirt is removed by pickling. Annealing treatment uses a higher temperature and a longer time to allow the zirconium alloy structure to fully recrystallize and form larger grains, thereby improving the creep resistance of the zirconium alloy tube.

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

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

[0047] In order to verify the existence of the precipitated 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, and the balance is Zr and impurities, and the impurities contain 0.015% Si. The transmission electron microscope photo of its structure is shown in FIG. Figure 1 As shown, the precipitate phase 1 can be seen. Electron diffraction analysis of the precipitate phase 1 is performed, and the results are as follows Figure 2 As shown, combined with Figure 3 The energy spectrum analysis shown can identify the precipitated phase 1 as the Zr(Nb,Fe,V,Cu)2 precipitated phase.

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

[0049] In some embodiments, 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 by weight. 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 a second phase and reduce anion vacancies to improve the strength and corrosion resistance of the alloy. In a specific embodiment, the element composition of the zirconium alloy is as follows by weight: 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, the zirconium alloy contains, by weight, 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, 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 by weight. In a specific embodiment, the element composition of the zirconium alloy 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, and the balance is Zr.

[0052] The compositions of the Zr alloys in a group of preferred embodiments and comparative examples are shown in Table 1 by weight.

[0053]

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

[0055] The corrosion weight gain curves of the embodiment and the comparative example are shown in Figure 1. Figure 4 It can be seen that the corrosion resistance of the embodiment is improved by 63.1% compared with the Zr-4 alloy.

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

[0057] At the same time, since the embodiment has higher Nb and Fe element contents and lower Cr and V element contents, and O / N is controlled to be above 20, the embodiment has better radiation resistance growth performance.

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

[0059] The purpose of the above embodiments is to further explain the present invention in detail 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 of the present invention, the optimization or equivalent replacement of the technical features involved, and the combination of implementation methods in different embodiments without conflict of structure and principle, all fall within the protection scope of the present invention.

Claims

1. A bending-resistant zirconium alloy suitable for a reactor, characterized in that: The composition comprises, by weight: not less than 97% Zr, 0.61%-1.4% Sn, 0.6%-1.4% Nb, 0.05%-0.4% Fe, 0.1%-0.2% O, 0.001%-0.005% N, and 0.001%-0.06% Ge; A group consisting of 0.001%-0.3% Cr and 0.01%-0.3% Mo and / or 0.01%-0.09% V and 0.01%-0.09% Cu; Among them, O / N ≥ 20; The structure of the bending-resistant zirconium alloy suitable for the reactor includes Zr(Nb, Fe, V, Cu)2 and / or Zr(Nb, Fe, Cr, Mo)2 precipitation phases.

2. The bending-resistant zirconium alloy suitable for a reactor according to claim 1, characterized in that: The bending-resistant zirconium alloy suitable for the reactor contains, by weight, 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, and 0.001%-0.09% Cu; And 0.001%-0.06% Ge.

3. The bending-resistant zirconium alloy suitable for a reactor according to claim 1, characterized in that: The bending-resistant zirconium alloy suitable for the reactor contains, by weight, 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, and 0.001%-0.09% Mo; as well as 0.001%-0.06%Ge.

4. The bending-resistant zirconium alloy suitable for a reactor according to claim 3, characterized in that: The bending-resistant zirconium alloy suitable for the reactor contains, by weight, 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.

5. A method for manufacturing a bending-resistant zirconium alloy suitable for a reactor, characterized in that: Used to manufacture a bending-resistant zirconium alloy suitable for a reactor as claimed in any one of claims 1 to 4, and comprising the following steps: Step a): providing alloy raw materials according to the composition of the bending-resistant zirconium alloy suitable for the reactor, wherein Zr is nuclear-grade sponge zirconium, and the other alloy components are nuclear-grade metal ingots, and non-metallic elements are present in the nuclear-grade metal ingots; using a vacuum consumable arc furnace to melt the alloy raw materials at least three times to obtain a zirconium alloy ingot; Step b): forging the zirconium alloy ingot at 800°C-1100°C and keeping it warm for 1h-4h, performing shot blasting and machining after cooling, and then heating it to the β phase region and keeping it warm for 5min-10min to perform quenching treatment at not less than 50K / s to obtain a quenched billet; Step c): hot extruding the quenched blank at 600-650° C., followed by 3-4 passes of cold rolling, with the deformation of each pass of cold rolling being 40%-80%; Annealing treatment is performed between adjacent cold rolling processes, the annealing temperature is 560℃-600℃, and the annealing time is 1h-4h; The obtained product of bending-resistant zirconium alloy suitable for reactor is obtained.

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