A zirconium alloy for nuclear reactors, its preparation method and application

Abstract

The present invention provides a zirconium alloy for nuclear reactors, its preparation method and application, relating to the technical field of nuclear materials. The zirconium alloy provided by the present invention, by weight percentage, has chemical components including: Sn 0.40 - 0.65%, Nb 0.12 - 0.25%, Fe 0.35 - 0.50%, Cr 0.15 - 0.20%, Cu 0 - 0.13%, O 0.08 - 0.16%, and the balance is Zr. The zirconium alloy provided by the present invention is a stress-relieved or partially recrystallized structure, and the Zr(FeCr)2 second-phase precipitates are mainly distributed in the α-Zr matrix. It not only has good mechanical properties but also maintains excellent corrosion resistance in an oxygen-rich high-temperature and high-pressure water environment. Compared with existing Zr alloys, the zirconium alloy provided by the present invention has a lower corrosion weight gain when corroded in an oxygen-rich water environment for up to 240 days, meeting the application requirements of small and micro nuclear reactors under special water chemistry conditions.

Description

Technical Field

[0001] The present invention relates to the technical field of nuclear materials, and particularly relates to a zirconium alloy for nuclear reactors, a preparation method thereof, and an application thereof. Background Art

[0002] The corrosion resistance of the water side of the nuclear fuel cladding and the residual plasticity after service are directly related to the economy, safety, and advancement of the reactor. For water-cooled nuclear reactors, the dissolved oxygen (DO) in the cooling medium has an important impact on the corrosion resistance of the zirconium alloy used for the fuel cladding. On the one hand, different from large commercial nuclear reactors, in order to simplify the system or save space, some reactor types of small water-cooled reactors do not adopt a hydrogenation deoxidation device, resulting in an increase in the DO content in the primary loop water. Such an oxygen-rich water chemical environment is bound to affect the corrosion resistance of the fuel cladding. On the other hand, after long-term operation, the porous water scale deposited on the surface of the fuel cladding will reduce the fluidity of the cooling water. In addition to causing the enrichment of element B, it may also cause an increase in the local DO concentration. Therefore, developing a new zirconium alloy suitable for the DO water environment is of great significance for the large-scale commercialization of small and micro nuclear reactors, while improving their economy and ensuring their safety.

[0003] Currently, new zirconium alloys are still being continuously developed at home and abroad to improve their corrosion resistance, hydrogen absorption performance, mechanical properties, irradiation growth resistance, and irradiation creep resistance. Among them, the corrosion resistance is the most critical and most likely to change among the five properties of zirconium alloys, because the addition of very trace alloying elements (ppm-level changes) can cause a large change in the corrosion resistance of zirconium alloys. From the development process of zirconium alloys, alloying is an effective method to improve the comprehensive properties of zirconium alloys. Based on this, three major alloy series, namely Zr-Sn, Zr-Nb, and Zr-Sn-Nb, have been developed. By adding alloying elements such as Fe, Cr, Ni, and Cu, zirconium alloy materials such as Zr-2, Zr-4, E110, M5, ZIRLO, and E635 that have been commercialized, as well as a series of new zirconium alloys represented by N36 / N45 (China National Nuclear Corporation), CZ (China General Nuclear Power Group Co., Ltd.), and SZA (State Power Investment Corporation) with application prospects, have been formed. However, these high-performance zirconium alloys are designed for the water chemical environment of conventional nuclear reactors. In the DO water environment, generally speaking, the corrosion resistance of zirconium alloys gradually decreases with the increase in the Nb content, which is exactly opposite to the corrosion law of the conventional water chemical environment. For this, one possible explanation is that Nb is more likely to be oxidized into Nb2O5 second-phase particles under DO conditions, causing local volume expansion and greater additional stress of the oxide film, and then introducing more crystal defects such as vacancies, and gradually evolving into microvoids or microcracks at the columnar crystal boundaries of the oxide film, thereby accelerating the destruction of the compactness of the oxide film and promoting the interfacial diffusion of erosive media, resulting in an accelerated corrosion reaction process. Another possible explanation is that the high-valence Nb5+ The ionic doping effect of ZrO2 is changed, introducing more oxygen vacancies, thereby increasing the bulk diffusion of the erosive medium in the ZrO2 crystal and the migration rate in the oxide film, resulting in an accelerated corrosion rate. Therefore, these new Zr alloys containing Nb may be difficult to meet the requirements of small and micro nuclear reactors for high burnup, long cycle, and safety in an oxygen-rich water quality environment.

[0004] For this reason, attempts have been made to develop new Zr alloys for small and micro water-cooled nuclear reactors based on Nb-free Zr-Sn alloys such as Zr-2 or Zr-4. However, although Zr-2 has always been used as the fuel cladding material for boiling water reactors, the Ni element in it may increase the hydrogen absorption fraction of the α-Zr matrix; while the composition of Zr-4 does not seem to be within the optimal range, reducing Sn and increasing the Fe or Fe+Cr content can significantly improve the corrosion resistance in a conventional water chemistry environment. In fact, the corrosion resistance of Zr-Sn alloys in a conventional water chemistry environment is also significantly inferior to that of Zr-Nb or Zr-Sn-Nb alloys. Therefore, simply optimizing the composition based on Zr-Sn alloys still has deficiencies. Whether a small amount of Nb element can be added, or whether the adverse effects of Nb element in the DO water environment can be effectively curbed by the combined addition of elements such as Sn, Fe, Cr, and Cu, is still unknown.

[0005] In summary, the existing commercial alloys and new Zr alloys for conventional water-cooled nuclear reactor environments cannot meet the requirements of the fuel cladding material for small and micro water-cooled nuclear reactors in the DO water environment to a certain extent. Summary of the Invention

[0006] The purpose of the present invention is to provide a Zr alloy for nuclear reactors, its preparation method and application. The Zr alloy provided by the present invention has excellent corrosion resistance in an oxygen-rich high-temperature and high-pressure water environment, while maintaining good mechanical properties.

[0007] In order to achieve the above invention purpose, the present invention provides the following technical solutions:

[0008] The present invention provides a Zr alloy for nuclear reactors. By weight percentage, the chemical composition includes: Sn 0.40 - 0.65%, Nb 0.12 - 0.25%, Fe 0.35 - 0.50%, Cr 0.15 - 0.20%, Cu 0 - 0.13%, O 0.08 - 0.16%, and the balance is Zr.

[0009] Preferably, the Zr alloy includes an α-Zr matrix and second-phase precipitation particles; the α-Zr matrix is a hexagonal close-packed structure; the second-phase precipitation particles include Zr(FeCr)2 phase, o-Zr3Fe phase, and t-Zr2Cu phase.

[0010] Preferably, the average grain size of the α-Zr matrix is between 1.02±1.12 and 7.57±3.47 μm, being a stress-relieved or partially recrystallized microstructure, and the recrystallization fraction is between 13% and 78%; the second-phase precipitated particles are spherical or near-spherical, and are dispersed or banded distributed in the intragranular and grain boundary regions of the α-Zr matrix, with an average size between 25 and 420 nm.

[0011] The present invention provides a method for preparing the zirconium alloy for nuclear reactors as described in the above technical solution, comprising the following steps:

[0012] Melting metal raw materials to obtain a zirconium alloy ingot; the composition of the zirconium alloy ingot is consistent with the chemical composition of the zirconium alloy as described in the above technical solution;

[0013] Hot pressing or forging the zirconium alloy ingot to obtain a blank;

[0014] Successively subjecting the blank to homogenization treatment, hot rolling, cold rolling and annealing treatment to obtain a zirconium alloy for nuclear reactors.

[0015] Preferably, the temperature of the hot pressing is 660-720 °C; the temperature of the forging is 920-1120 °C.

[0016] Preferably, the temperature of the homogenization treatment is 1000-1070 °C, the holding time is 0.5-4 h, and then quenching to room temperature.

[0017] Preferably, the temperature of the hot rolling is 620-750 °C.

[0018] Preferably, the cold rolling is multi-pass rolling, the reduction per pass of the cold rolling is 18-45%, the intermediate annealing temperature between passes is 480-600 °C, and the holding time is 1-6 h.

[0019] Preferably, the temperature of the annealing treatment is 450-600 °C, the holding time is 3-7 h; the atmosphere of the annealing treatment is vacuum.

[0020] The present invention provides the application of the zirconium alloy for nuclear reactors as described in the above technical solution or the zirconium alloy for nuclear reactors prepared by the preparation method as described in the above technical solution in a small and micro water-cooled nuclear reactor.

[0021] The present invention provides a zirconium alloy for nuclear reactors, and by weight percentage, the chemical composition comprises: Sn 0.40-0.65%, Nb 0.12-0.25%, Fe 0.35-0.50%, Cr 0.15-0.20%, Cu 0-0.13%, O 0.08-0.16%, and the balance is Zr. In the present invention, the functions and composition selection bases of various additive elements are as follows:

[0022] Sn can increase the phase transformation point of zirconium alloys, facilitating processing and deformation in the α-Zr single-phase region at higher temperatures, and reducing or avoiding the residual high-temperature β-Zr phase in the room-temperature microstructure. Meanwhile, due to the segregation of solute Sn at defects such as grain boundaries in the oxide film, Sn can counteract the harmful effects of impurity elements such as N, C, and Al in sponge zircon on corrosion resistance; due to solid solution strengthening in α-Zr, Sn can improve the mechanical properties of zirconium alloys such as yield strength, tensile strength, and creep resistance; in addition, Sn can balance the nodular corrosion resistance and irradiation creep resistance. However, excessive Sn may increase the volume fraction of tetragonal t-ZrO2 during corrosion, causing greater additional stress during its transformation to monoclinic m-ZrO2 and damaging the compactness of the oxide film, thereby inducing corrosion transition and reducing corrosion resistance. Currently, high-performance nuclear-grade zirconium alloys generally adopt a low-Sn composition design. Therefore, the Sn content is controlled at a relatively low level (lower than that of commercial alloys such as Zr-4 and ZIRLO).

[0023] Nb can dissolve in the α-Zr matrix or form β-Nb second-phase precipitated particles, but the maximum solid solubility of Nb in α-Zr is affected by other alloying elements. In the primary loop water environment of a conventional pressurized water reactor, Nb can significantly improve the corrosion resistance of the alloy, while in an oxygen-rich water or high-temperature steam environment, Nb may deteriorate the corrosion resistance. Therefore, the Nb content is selected to be lower than the solid solubility content in α-Zr to avoid the adverse effects caused by the delayed oxidation of β-Nb particles in the oxide film. Additionally, since Nb can improve the irradiation growth resistance and other properties of zirconium alloys, the addition of Nb is not completely abandoned, but the addition of other alloying elements such as Fe and Cr is used to form SPPs to minimize the damage of solid solution Nb to corrosion resistance.

[0024] The addition of Fe can first prevent the decline in mechanical properties caused by the reduction of Sn and Nb contents in the alloy, and secondly promote the formation of the Zr(Fe,Cr)2 second phase to improve the hydrogen absorption performance of the alloy matrix, nodular corrosion resistance, and irradiation growth resistance. However, excessive Fe may form a Zr3Fe phase with a large size that is not conducive to corrosion resistance.

[0025] The role of Cr is similar to that of Fe. It can improve the comprehensive mechanical properties, form the Zr(Fe,Cr)2 second phase in cooperation with Fe, and improve the nodular corrosion resistance, irradiation growth resistance, and creep resistance of the alloy, especially improving the corrosion resistance of the cladding-end plug weld. However, excessive Cr is not conducive to the processing of zirconium alloys and increases the hydrogen absorption fraction of the alloy.

[0026] O has a high solid solubility in α-Zr, which can be as high as 29at.%. Therefore, on the one hand, O can affect the thickness and morphology of the hexagonal h-ZrO transition layer and the O supersaturated solid solution layer near the oxide film / matrix (O / M) interface during corrosion, thereby affecting the oxidation reaction process; on the other hand, O can change the strength of the metal matrix at the oxidation front through solid solution strengthening, affect the degree of stress relaxation in the oxide film, and thus affect the corrosion transition. In addition, O can improve the alloy's resistance to radiation growth and creep resistance. However, too much O is not conducive to the processing of zirconium alloys.

[0027] Cu diffuses more slowly than Fe in α-Zr, and tends to replace Fe in the Zr(Fe,Cr)2 phase and enrich it, thus changing the kinetics of the precipitation reaction and making the Zr(Fe,Cr)2 phase finer and more dispersed; trace amounts of Cu can also inhibit the growth and excessive coarsening of the Zr3Fe phase, preventing it from generating greater additional stress during oxidation. However, excessive Cu may form a coarser Zr2Cu phase, which will introduce greater local stress into the oxide film and destroy its density, thereby deteriorating the corrosion resistance. In addition, Cu lowers the recrystallization temperature of the alloy, broadens the alloy processing window, increases the recrystallization fraction of α-Zr during the final annealing, and reduces its equivalent grain size. In addition, Cu in the solid solution state can inhibit the high-valence Nb 5+ Formation slows down the corrosion reaction rate in oxygen-rich water environment.

[0028] In addition to the single effects of the above alloying elements, the coupling effects of multiple elements can be produced in actual alloys, and beneficial effects can be produced. For example, Nb is completely in solid solution below the limit solid solubility. When Fe and Cr exist and form Zr(Fe,Cr)2 phase, Nb can diffuse and enrich in it, further reducing its solid solution content; at the same time, when Cu exists, Nb and Cu have lower diffusion rates in α-Zr than Fe, and both can inhibit the growth of Zr(Fe,Cr)2 phase, thereby obtaining a fine and dispersed SPPs distribution. In addition, even if there is a very small amount of solid-dissolved Nb in α-Zr, Cu can also weaken its adverse effects. It can be seen that as long as the composition interval is reasonably selected, the coupling addition of multiple elements can not only optimize microstructures such as SPPs, but also inhibit the adverse effects of certain elements (such as Nb in oxygen-rich water environment), thereby achieving the purpose of simultaneously improving mechanical properties and improving corrosion resistance, and expanding the application of zirconium alloys in special water chemistry.

[0029] Combining the single and coupled effects of the above various elements and fully considering the application environment of oxygen-rich high-temperature and high-pressure water, the zirconium alloy provided by the present invention adopts a composition design of "medium Sn and low Nb" and contains an appropriate amount of Cu, so that a Zr-Sn-Fe-Cr series alloy has excellent corrosion resistance in an oxygen-rich water environment while maintaining good mechanical properties, meeting the requirements of small and micro water-cooled nuclear reactors for improving economy and safety.

[0030] The results of the examples show that the zirconium alloy provided by the present invention has excellent corrosion resistance in an oxygen-rich high-temperature and high-pressure water environment. Compared with the existing commercial M5, ZIRLO, and Zr-4 alloys, it has a lower corrosion weight gain after 240 days of corrosion in two oxygen-containing water environments of 300 ppb and 1000 ppb, and is suitable for use as fuel cladding, grid strips, and structural components of small and micro nuclear reactors. Description of the Drawings

[0031] Figure 1 It is a grain morphology map and size distribution map of the electron backscatter diffraction (EBSD) of the zirconium alloys prepared in Examples 1-2 and Comparative Examples 1-2;

[0032] Figure 2 It is a microstructure morphology map of the scanning electron microscope (SEM) and transmission electron microscope (TEM) of the zirconium alloys prepared in Examples 1-2 and Comparative Examples 1-2;

[0033] Figure 3 It is a transmission electron microscope (TEM) morphology of the second phase in the zirconium alloys prepared in Example 1 and Example 3, as well as the corresponding selected area electron diffraction (SAED) pattern and energy spectrum (EDS) composition analysis diagram;

[0034] Figure 4 It is a comparison diagram of the corrosion weight gain curves of the zirconium alloys prepared in Examples 1-2 and Comparative Examples 1-2 and the reference commercial alloys in a 300 ppb dissolved oxygen high-temperature and high-pressure water environment (360 °C, 20 MPa);

[0035] Figure 5 It is a comparison diagram of the corrosion weight gain curves of the zirconium alloys prepared in Examples 3-4 and Comparative Examples 3-4 and the reference commercial alloys in a 1000 ppb dissolved oxygen high-temperature and high-pressure water environment (360 °C, 20 MPa);

[0036] Figure 6 It is a comparison diagram of the corrosion weight gain curves of the zirconium alloys prepared in Examples 4-5 and Comparative Example 5 and the reference commercial alloys in a 1000 ppb dissolved oxygen high-temperature and high-pressure water environment (360 °C, 20 MPa);

[0037] Figure 7Scanning electron microscope (SEM) cross-sectional fracture images (300 ppb) of the oxide films of the 0.07Cu zirconium alloys prepared in Examples 2, 4, 5 and Comparative Example 5 after 240 days of corrosion in high-temperature and high-pressure water, and energy dispersive spectroscopy (EDS) chemical composition surface distribution maps and crystal structure analysis maps (1000 ppb) of the oxide films by transmission electron microscope (TEM). Detailed implementation manners

[0038] The present invention provides a zirconium alloy for nuclear reactors. By weight percentage, the chemical composition includes: Sn 0.40 - 0.65%, Nb 0.12 - 0.25%, Fe 0.35 - 0.50%, Cr 0.15 - 0.20%, Cu 0 - 0.13%, O 0.08 - 0.16%, and the balance is Zr.

[0039] By weight percentage, the zirconium alloy for nuclear reactors provided by the present invention includes Sn 0.40 - 0.65%, preferably 0.45 - 0.50%. In the present invention, Sn can increase the phase transformation point of the zirconium alloy, facilitate processing and deformation in the α-Zr single-phase region at higher temperatures, and reduce or avoid the residual high-temperature β-Zr phase in the room-temperature structure. At the same time, due to the segregation of solid-solution Sn at defects such as grain boundaries of the oxide film, Sn can offset the harmful effects of impurity elements such as N, C, and Al in sponge zircon on the corrosion resistance; due to solid-solution strengthening in α-Zr, Sn can improve the mechanical properties of the zirconium alloy such as yield strength, tensile strength, and creep resistance; in addition, Sn can balance the nodular corrosion resistance and irradiation creep resistance. However, excessive Sn may increase the volume fraction of tetragonal structure t-ZrO2 during the corrosion process, and when it transforms to monoclinic structure m-ZrO2, it causes greater additional stress, destroys the denseness of the oxide film, thereby inducing corrosion transition and reducing the corrosion resistance. Currently, high-performance nuclear zirconium alloys generally adopt a low-Sn composition design. Therefore, the Sn content is controlled at a relatively low level (lower than commercial alloys such as Zr-4 and ZIRLO).

[0040] By weight percentage, the zirconium alloy for nuclear reactors provided by the present invention comprises 0.12 to 0.25% of Nb, preferably 0.15 to 0.20%. In the present invention, Nb can be dissolved in the α-Zr matrix or form β-Nb second-phase precipitation particles, but the maximum solid solubility of Nb in α-Zr is affected by other alloying elements. In the primary loop water environment of a conventional pressurized water reactor, Nb can significantly improve the corrosion resistance of the alloy, while in an oxygen-rich water or high-temperature steam environment, Nb may deteriorate the corrosion resistance. Therefore, the content of Nb is selected to be lower than the solid solubility content in α-Zr to avoid the adverse effects caused by the delayed oxidation of β-Nb particles in the oxide film. In addition, since Nb can improve the properties such as radiation growth resistance of the zirconium alloy, the addition of Nb is not completely abandoned, but the damage of solid solution Nb to the corrosion resistance is minimized by the addition of other alloying elements such as Fe and Cr to form SPPs.

[0041] By weight percentage, the zirconium alloy for nuclear reactors provided by the present invention comprises 0.35 to 0.50% of Fe, preferably 0.36 to 0.49%. In the present invention, the addition of Fe can first avoid the decline in mechanical properties caused by the reduction of the contents of Sn and Nb in the alloy, and secondly can promote the formation of the Zr(Fe,Cr)2 second phase to improve the matrix hydrogen absorption performance, pitting corrosion resistance and radiation growth resistance of the alloy. However, excessive Fe may form a Zr3Fe phase with a relatively large size that is not conducive to the corrosion resistance.

[0042] By weight percentage, the zirconium alloy for nuclear reactors provided by the present invention comprises 0.15 to 0.20% of Cr, preferably 0.18 to 0.19%. In the present invention, the role of Cr is similar to that of Fe. It can improve the comprehensive mechanical properties, and at the same time cooperate with Fe to form the Zr(Fe,Cr)2 second phase, and improve the pitting corrosion resistance, radiation growth resistance and creep resistance of the alloy, especially improving the corrosion resistance of the cladding-end plug weld. However, too much Cr is not conducive to the processing of the zirconium alloy and also increases the hydrogen absorption fraction of the alloy.

[0043] By weight percentage, the zirconium alloy for nuclear reactors provided by the present invention comprises 0 to 0.13% of Cu, preferably 0.02 to 0.10%, and more preferably 0.05 to 0.07%. In the present invention, Cu has a slower diffusion rate in α-Zr than Fe, and tends to replace Fe in the Zr(Fe,Cr)2 phase and enrich therein, thus changing the kinetics of the precipitation reaction and making the Zr(Fe,Cr)2 phase finer and more dispersed; trace amounts of Cu can also inhibit the growth and excessive coarsening of the Zr3Fe phase, avoiding greater additional stress during oxidation. However, excessive Cu may form coarser Zr2Cu phases, which damage the compactness by introducing greater local stress in the oxide film, thereby deteriorating the corrosion resistance. In addition, Cu reduces the recrystallization temperature of the alloy, broadens the alloy processing window, increases the recrystallization fraction of α-Zr during final annealing, and reduces its equivalent grain size. In addition, solid-solution Cu can inhibit the formation of high-valent Nb 5+ from forming and slow down the corrosion reaction rate in an oxygen-rich water environment. By controlling the content of Cu within the above range, the present invention can obtain a suitable α-Zr matrix grain size and recrystallization fraction, and obtain a finer and more dispersed Zr(FeCr)2 phase, avoiding the large-scale generation of larger-sized Zr3Fe and Zr2Cu phases.

[0044] By weight percentage, the zirconium alloy for nuclear reactors provided by the present invention comprises 0.08 to 0.16% of O, preferably 0.12 to 0.13%. In the present invention, O has a high solubility in α-Zr, up to 29 at.%. Therefore, on the one hand, O can affect the thickness and morphology of the hexagonal h-ZrO transition layer and the O supersaturated solid solution layer near the oxide film / matrix (O / M) interface during the corrosion process, thereby affecting the oxidation reaction process; on the other hand, O can change the strength of the metal matrix at the oxidation front through solid solution strengthening, affecting the stress relaxation degree in the oxide film, and thus affecting the corrosion transition. In addition, O can improve the irradiation growth resistance and creep resistance of the alloy. However, too much O is not conducive to the processing of zirconium alloys.

[0045] The zirconium alloy for nuclear reactors provided by the present invention comprises the balance of Zr, specifically nuclear-grade sponge zirconium with Hf removed. In a specific embodiment of the present invention, the zirconium alloy further comprises inevitable impurities, and the added alloying elements can avoid their adverse effects.

[0046] The present invention makes full use of the single and coupled effects of the above alloying elements and produces beneficial effects. For example, when Nb is below the limit solid solubility, it is completely in a solid solution state. When Fe and Cr are present and form the Zr(Fe,Cr)2 phase, Nb can diffuse and enrich therein, further reducing its solid solution content. At the same time, when Cu is present, since the diffusion rates of Nb and Cu in α-Zr are both lower than that of Fe, both can inhibit the growth of the Zr(Fe,Cr)2 phase, thereby obtaining a fine and dispersed SPPs distribution. In addition, even if there is a very small amount of solid solution Nb in α-Zr, Cu can weaken its adverse effects. It can be seen that as long as the composition range is reasonably selected, the coupled addition of multiple elements can not only optimize the microstructure such as SPPs, but also inhibit the adverse effects of certain elements (such as Nb in an oxygen-rich water environment), thereby achieving the purpose of simultaneously improving mechanical properties and corrosion resistance, and expanding the application of zirconium alloys in special water chemistries.

[0047] In the present invention, the microstructure of the zirconium alloy preferably includes an α-Zr matrix and second-phase precipitated particles (SPPs).

[0048] In the present invention, the α-Zr matrix is a stress-relieved or partially recrystallized microstructure, with a recrystallization fraction between 13% and 78%, and an average grain size between 1.02±1.12 μm and 7.57±3.47 μm. The α-Zr matrix preferably has a hexagonal close-packed structure (hcp), and there is no β-Zr with a body-centered cubic structure (bcc).

[0049] In the present invention, the second-phase precipitated particles preferably include the hcp- and face-centered cubic structure (fcc) Zr(FeCr)2 phase, as well as the possibly existing base-centered orthorhombic o-Zr3Fe phase and tetragonal t-Zr2Cu phase. Among them, the hcp- / fcc-Zr(FeCr)2 phase is the main one, and its volume fraction is preferably 78-95 vol.%, showing a spherical or quasi-spherical shape, and being dispersed or banded and distributed in the intragranular and intergranular regions of the α-Zr matrix, with an average size between 25 nm and 420 nm, preferably 25 nm to 120 nm.

[0050] In the present invention, the Cu element plays an important role in the microstructure. The addition of trace Cu increases the average grain size and recrystallization fraction of the α-Zr matrix, and changes the distribution, type, size and composition of the SPPs. When no Cu is added, the SPPs are mainly large-sized o-Zr3Fe phase and a small amount of fine fcc-Zr(FeCr)2 phase; when trace Cu is added, Cu replaces part of Fe and enriches in the Zr(FeCr)2 phase, thereby affecting its precipitation kinetics and refining its size. In addition, there are also a small amount of Cu-rich o-Zr3Fe phase and Fe-rich t-Zr2Cu phase; when excessive Cu is added, large-sized t-Zr2Cu phase is formed, resulting in an increase in the average size of the second phase. Therefore, the content of Cu is controlled to optimize the microstructure of the alloy.

[0051] The present invention provides a method for preparing the zirconium alloy for nuclear reactor as described in the above technical solution, comprising the following steps:

[0052] Melting metal raw materials to obtain a zirconium alloy ingot; the composition of the zirconium alloy ingot is consistent with the chemical composition of the zirconium alloy described in the above technical solution;

[0053] Hot pressing or forging the zirconium alloy ingot to obtain a blank;

[0054] Successively subjecting the blank to homogenization treatment, hot rolling, cold rolling and annealing treatment to obtain a zirconium alloy for nuclear reactor.

[0055] In the present invention, melting metal raw materials to obtain a zirconium alloy ingot; the composition of the zirconium alloy ingot is consistent with the chemical composition of the zirconium alloy described in the above technical solution. In the present invention, the metal raw materials preferably include nuclear-grade sponge zirconium and pure metal, or nuclear-grade sponge zirconium and master alloy. In the present invention, the melting is preferably vacuum consumable arc melting. In the present invention, the temperature of the melting is preferably 1250-1885 °C, the time is preferably 30-60 minutes, and the number of flips is 3-7 times. The present invention preferably air-cools to room temperature after melting to obtain a zirconium alloy ingot.

[0056] After obtaining the zirconium alloy ingot, the present invention hot presses or forges the zirconium alloy ingot to obtain a blank. In the present invention, the temperature of the hot pressing is preferably 660-720 °C, more preferably 680-700 °C. In the present invention, the composition of the hot pressing is relatively uniform, without casting defects, and the deformation temperature and load are relatively low, which is suitable for laboratory small-sized ingots (≤200 g).

[0057] In the present invention, the temperature of the forging is preferably 920-1120 °C, more preferably 950-1000 °C. In the present invention, the forging is preferably applicable to industrial-grade large and medium-sized ingots.

[0058] Preferably after the hot pressing or forging, the obtained material is successively descaled, pickled and washed with water to obtain a blank. In the present invention, the pickling solution preferably used is a mixed acid; the mixed acid preferably includes hydrofluoric acid, nitric acid solution, sulfuric acid solution and water. In the present invention, the mass concentration of the hydrofluoric acid is preferably >40%; the mass concentration of the nitric acid solution is preferably 65-68%; the mass concentration of the sulfuric acid solution is preferably 95-98%. In the present invention, the water is preferably deionized water. In the present invention, the volume ratio of the hydrofluoric acid, nitric acid solution, sulfuric acid solution and water is preferably 1:3:3:3. In the present invention, the water washing is preferably washing with deionized water. In the present invention, the surface of the blank is bright and has a metallic luster.

[0059] After obtaining the blank, the present invention successively subjects the blank to homogenization treatment, hot rolling, cold rolling and annealing treatment to obtain a zirconium alloy for nuclear reactors. In the present invention, the temperature of the homogenization treatment is preferably 1000-1070 °C, more preferably 1050 °C, in the β-Zr single-phase region; the holding time of the homogenization treatment is preferably 0.5-4 h, more preferably 1 h. The present invention preferably quenches to room temperature after the homogenization treatment, and the cooling rate of quenching is preferably >30 °C / s, more preferably >100 °C / s. In the present invention, the quenching method is preferably water cooling. After β-phase quenching in the present invention, alloying elements are fully dissolved, and during the subsequent annealing treatment, alloying elements such as Fe and Cr are precipitated dispersedly in the form of SPPs, and only a very small amount is dissolved in the α-Zr matrix.

[0060] In the present invention, the temperature of the hot rolling is preferably 620-750 °C, more preferably 680 °C. In the present invention, the hot rolling is preferably multi-pass hot rolling, and the interval time between passes is preferably 5-12 min; the number of passes of the hot rolling is preferably 4-7 passes, more preferably 5 passes; the reduction per pass of the hot rolling is preferably 12-55%; the total reduction of the hot rolling is preferably 70-90%. The present invention preferably air-cools the obtained hot-rolled sheet to room temperature after the hot rolling.

[0061] In the present invention, the cold rolling is preferably multi-pass rolling, more preferably 4-6 passes; the reduction per pass of the cold rolling is preferably 18-45%, more preferably 20%; the total reduction of the cold rolling is preferably 75-95%. In the present invention, between passes of the cold rolling, it is preferably vacuum annealed at 480-600 °C for 1-6 h, more preferably vacuum annealed at 580 °C for 2 h. In the present invention, the cumulative annealing parameter A value (which mainly affects the final alloy microstructure and SPPs distribution characteristics) of the vacuum annealing is 10 -18 h; the A = ∑t i ×exp(-Q / RT i), Q / R = 40000K; T is the annealing temperature in K; t is the annealing time in h; Q is the activation energy.

[0062] According to different rolling methods, the present invention obtains zirconium alloy plates or tubes with predetermined sizes and specifications.

[0063] In the present invention, the temperature of the annealing treatment is preferably 450 - 600 °C, more preferably 470 - 550 °C; the holding time is preferably 3 - 7 h, more preferably 5 h. In the present invention, the atmosphere of the annealing treatment is preferably vacuum. The present invention preferably air-cools to room temperature after the annealing treatment.

[0064] In the present invention, when the temperature of the annealing treatment is 550 °C, the highest recrystallization fraction of the α-Zr matrix is 77.2%, which is a partially recrystallized structure composed of equiaxed grains and lath-shaped grains; when the temperature of the annealing treatment is 470 °C, the recrystallization fraction is 13%, and the α-Zr matrix is considered to be a stress-relieving annealing structure dominated by lath-shaped grains.

[0065] The yield strength of the partially recrystallized zirconium alloy (annealing temperature is 550 °C or 510 °C) provided by the present invention is 390 ± 12 - 425 ± 13 MPa, the tensile strength is 542 ± 9 - 558 ± 8 MPa, and the elongation is 22 ± 1 - 24 ± 2%; the yield strength of the stress-relieved zirconium alloy (annealing temperature is 470 °C) is 565 ± 9 - 589 ± 10 MPa, the tensile strength is 772 ± 6 - 792 ± 6 MPa, and the elongation is 16 ± 1 - 18 ± 1%. This mechanical property meets the material selection requirements of zirconium alloys for nuclear reactor fuel cladding.

[0066] The present invention also provides the application of the zirconium alloy for nuclear reactors described in the above technical solution or the zirconium alloy for nuclear reactors obtained by the preparation method described in the above technical solution in small and micro water-cooled nuclear reactors, preferably used as fuel cladding materials, more preferably used for the positioning grids or plate-like members of tubular fuel claddings or fuel cladding tubes. In the present invention, to evaluate the applicability of the zirconium alloy in small and micro water-cooled nuclear reactors, the experimental environment is preferably oxygen-rich high-temperature and high-pressure water; the dissolved oxygen content of the oxygen-rich water quality is preferably 300 - 1000 ppb to simulate a harsh strongly oxidizing environment and accelerate the evaluation process; the temperature of the high-temperature and high-pressure water environment is preferably 360 °C, and the pressure is preferably 20 MPa.

[0067] The low-Nb Cu-containing zirconium alloy provided by the present invention has equivalent uniform corrosion resistance to Zircaloy alloys such as Zr-2 or Zr-4 without Nb in a high-temperature and high-pressure water environment with a dissolved oxygen content of 300-1000 ppb, and is superior to commercial Zr-Nb-based M5 and Zr-Sn-Nb-based ZIRLO alloys. Moreover, there is no tendency for nodular corrosion to occur after 240 days of corrosion, making it suitable for the oxygen-rich water environment of small and micro water-cooled nuclear reactors.

[0068] Next, the technical solutions of the present invention will be clearly and completely described in conjunction with the embodiments of the present invention. Obviously, the described embodiments are only a part of the embodiments of the present invention, rather than all of the embodiments. All other embodiments obtained by those of ordinary skill in the art based on the embodiments of the present invention without creative efforts shall fall within the protection scope of the present invention.

[0069] Examples 1-5 and Comparative Examples 1-5

[0070] First step: Alloy melting

[0071] According to the chemical composition of the alloy in Table 1, nuclear-grade sponge zirconium, Sn pure metal, Zr-Cu, Zr-Fe, Zr-Cr, and Zr-Nb master alloy raw materials are melted into a zirconium alloy ingot with a weight of 200 g using a vacuum consumable arc furnace, and then hot-pressed into a disc shape at 680 °C. After removing the surface oxide scale, pickling and water washing are carried out in sequence to obtain a billet with a bright surface.

[0072] Second step: Rolling and heat treatment

[0073] The above billet is heated to 1050 °C and held for 1 h for homogenization treatment, and then water-quenched to room temperature; then hot-rolled in 5 passes at 680 °C with a 5-min interval between passes, and air-cooled to room temperature; then subjected to 6 passes of cold rolling + intermediate annealing treatment, with a reduction per cold rolling pass of 20%, and vacuum annealing at 580 °C for 2 h between passes. The cumulative annealing parameter A value is 1.02×10 -18 h, and finally a zirconium alloy sheet with a thickness of 0.6 mm is obtained to obtain the formed alloy.

[0074] Third step: Final annealing

[0075] The formed alloy is heated to the final annealing temperature shown in Table 1, held for 5 h for vacuum annealing, and air-cooled to room temperature to obtain the zirconium alloy.

[0076] Table 1 Alloy compositions, annealing temperatures, and dissolved oxygen concentrations of Examples 1-5 and Comparative Examples 1-5

[0077]

[0078] Test Example 1

[0079] Figure 1 It is the grain morphology map and the corresponding size distribution map of the electron backscatter diffraction (EBSD) of different Cu-content zirconium alloy matrices during annealing at 550 °C for 5 h in Examples 1-2 and Comparative Examples 1-2. From Figure 1 It can be seen that when no Cu is added (Comparative Example 2), there are obvious lath-shaped non-recrystallized structures in the zirconium alloy. At this time, the recrystallization fraction is 55.9%, and the equivalent grain size is 1.14 ± 1.12 μm; with the addition of Cu (Example 1) and the increase in content (Examples 2 and Comparative Example 1), the recrystallization fraction increases from 61.4% to 77.2%, and at the same time, the grain size increases from 1.53 ± 1.96 μm to 7.57 ± 3.47 μm. It can be seen that under the same preparation process and final annealing conditions, both the recrystallization degree and the average grain size increase with the increase in Cu content, that is, the addition of Cu is beneficial to obtaining equiaxed grains with larger sizes. At the same time, there are certain orientation differences among these grains. When the zirconium alloy is corroded, the higher the grain boundary ratio of the α-Zr matrix, the greater the crystal orientation difference, and the easier it is to promote the corrosion reaction. Therefore, the addition of an appropriate amount of Cu improves the corrosion resistance of the zirconium alloy by increasing the equivalent grain size, reducing the grain boundary ratio and orientation difference.

[0080] Test Example 2

[0081] Figure 2Scanning electron microscopy (SEM) images and transmission electron microscopy (TEM) images of the precipitation of second-phase particles (SPPs) in zirconium alloys with different Cu contents during annealing at 550 °C for 5 h in Examples 1-2 and Comparative Examples 1-2. From the SEM morphology, the SPP size distributions of the 0Cu and 0.05Cu zirconium alloys are similar, being fine and dispersed, while in the 0.07Cu and 0.13Cu zirconium alloys, the sizes of the SPPs are significantly increased and the number density is significantly decreased, mainly distributed at the grain boundaries. From the TEM morphology, the grain morphology distribution is consistent with the EBSD results in Test Example 1, that is, the recrystallization degree of the 0Cu zirconium alloy is the lowest, with a large number of defects such as dislocations distributed, and with the increase of the Cu content, the grain size increases and the defects such as dislocations decrease. From the TEM morphology, there are more SPPs in the 0Cu zirconium alloy, with irregular distribution and large size differences, which may indicate the existence of different types of SPPs; when 0.05 wt.% Cu is added, the precipitation of large-sized SPPs is inhibited, and the number is significantly reduced, and only relatively fine SPPs appear in the grains; with the increase of the Cu content to 0.07 wt.% and 0.13 wt.%, fine and dispersed SPPs are formed in the grains, and the SPPs at individual grain boundaries are relatively large. It can be seen that trace Cu can inhibit the coarsening of SPPs in the Zr-Sn-Nb-Fe-Cr alloy and improve the distribution morphology of SPPs, making them finer and more dispersed (compared with the Cu-free zirconium alloy). These differences in the size and distribution of SPPs may affect their characteristics of delaying oxidation in the oxide film: on the one hand, when maintaining the metal properties, the stress distribution in the oxide film is improved by deformation to reduce stress concentration, and on the other hand, during final oxidation, due to a higher Pilling-Bedworth (P-B) ratio, volume expansion generates additional stress on the surrounding oxide film, thereby affecting the defects and integrity in the oxide film. In summary, the addition of an appropriate amount of Cu promotes the precipitation of SPPs to be finer and more dispersed, thereby improving the corrosion resistance of the zirconium alloy.

[0082] Test Example 3

[0083] Figure 3 Transmission electron microscopy (TEM) morphology of the second phase, corresponding selected area electron diffraction (SAED) patterns, and energy-dispersive spectroscopy (EDS) composition analysis patterns for the zirconium alloys prepared in Example 1 and Example 3. Figure 3 (a) shows the TEM bright-field phase morphology and EDS composition of the precipitated phase SPP1, and (a1) is the corresponding SAED pattern of SPP1, calibrated as (b) shows the TEM bright-field phase morphology and EDS composition of the precipitated phase SPP2, and (b1) is the corresponding SAED pattern of SPP2, calibrated as (c) shows the TEM bright-field phase morphology and EDS composition of the precipitated phase SPP3, and (c1) is the corresponding SAED pattern of SPP3, calibrated as (d) is the TEM bright-field image morphology and EDS composition of the precipitation phase SPP4, and (d1) is the SAED pattern of the corresponding SPP4, calibrated as (e) is the TEM bright-field image morphology and EDS composition of the precipitation phase SPP5, and (e1) is the SAED pattern of the corresponding SPP5, calibrated as (f) is the TEM bright-field image morphology and EDS composition of the precipitation phase SPP6, and (f1) is the SAED pattern of the corresponding SPP6, calibrated as It can be seen that in the alloy containing 0.05Cu, there are three types of secondary phase precipitations, namely Zr(FeCr)2 phase, Cu-rich Zr3Fe phase and Fe-rich Zr2Cu phase. SAED shows that the Zr(FeCr)2 phase has two structures, hexagonal close-packed (hcp) and face-centered cubic (fcc), the Zr3Fe phase is a base-centered orthorhombic structure (o), and the Zr2Cu phase is a tetragonal structure (t). EDS analysis shows that the Zr(FeCr)2 phase contains a certain amount of Nb, the Zr3Fe phase contains Cu, and the Zr2Cu phase contains Fe. It can be seen that the addition of trace Cu has an important influence on the distribution characteristics of SPPs. The addition of trace Cu replaces part of Fe and enriches in the Zr(FeCr)2 phase, thereby affecting its precipitation kinetics, refining its size, and then improving the corrosion resistance of the alloy. In addition, when the Cu content is too low (<0.02 wt.%), a coarse-grained o-Zr3Fe phase appears; when the Cu content is too high (>0.08 wt.%), a larger-sized t-Zr2Cu phase is formed. These secondary phases may all reduce the corrosion resistance of the zirconium alloy, so the optimal content of Cu is between 0.02 and 0.08 wt.%.

[0084] Test Example 4

[0085] Figure 4 are the corrosion weight gain curves of the zirconium alloys prepared in Examples 1-2 and Comparative Examples 1-2 and the reference commercial alloy in a high-temperature water environment with 300 ppb dissolved oxygen (360 °C, 20 MPa). From Figure 4It can be seen that the 0.07Cu zirconium alloy (Example 2) has the best corrosion resistance, with a slow initial corrosion rate and the lowest weight gain after 240 days. The corrosion pattern of the 0.05Cu zirconium alloy (Example 1) is similar to that of the 0.07Cu alloy, but the weight gain rate is significantly higher than that of the 0.05Cu zirconium alloy around the corrosion transition (near 60 - 90 days), so its weight gain after 240 days is higher. Judging from the weight gain after 240 days, the corrosion resistance of the alloys involved in Examples 1 - 2 is better than that of the commercial Zr-4 alloy. The 0.13Cu zirconium alloy (Comparative Example 1) has a fast initial corrosion rate, comparable to the reference Zr-4, M5, and ZIRLO alloys. Even though the corrosion transition time and the corrosion rate after the transition are not much different from those of the 0.05Cu and 0.07Cu zirconium alloys, it still results in a much higher final corrosion weight gain (after 240 days) than the zirconium alloys in Examples 1 and 2, second only to the M5 alloy. The 0Cu zirconium alloy (Comparative Example 2) has the lowest initial corrosion rate, but its corrosion rate increases rapidly after the transition, so its final weight gain is also relatively high, slightly lower than that of the 0.13Cu zirconium alloy. The corrosion resistance (weight gain after 240 days) of the alloys involved in Comparative Examples 1 - 2 is inferior to that of the Zr-4 alloy. For the 0.13Cu zirconium alloy, it is due to its high initial corrosion rate, and for the 0Cu zirconium alloy, it is due to its high corrosion rate after the transition. In summary, due to the low initial corrosion rate and the corrosion rate after the transition, the 0.05Cu and 0.07Cu zirconium alloys have better corrosion resistance than other comparative alloys and reference commercial alloys in a 300 ppb dissolved oxygen water environment, so they are suitable for oxygen-rich high-temperature and high-pressure water environments.

[0086] Test Example 5

[0087] Figure 5 Corrosion weight gain curves of the zirconium alloys prepared for Examples 3 - 4 and Comparative Examples 3 - 4, as well as the reference commercial alloy, in a 1000 ppb dissolved oxygen high-temperature water environment (360 °C, 20 MPa). From Figure 5It can be seen that the 0.05Cu and 0.07Cu zirconium alloys (Examples 3 and 4) have similar corrosion laws. Although their initial corrosion rates are significantly increased compared to the 300 ppb dissolved oxygen environment and are comparable to those of the Zr-4 alloy, their corrosion rates are slower after the turning point, and the final weight gain (240 d) is only slightly higher than that of the Zr-4 alloy. Considering that it is difficult to reach 1000 ppb of dissolved oxygen content in the actual small and micro reactor conditions and comprehensively considering the corrosion resistance in the 300 ppb environment, it is considered that the 0.05Cu and 0.07Cu zirconium alloys are alternative alloys more suitable for the oxygen-rich water environment of small and micro nuclear reactors. However, the 0.13Cu zirconium alloy (Comparative Example 3) has good corrosion resistance due to its low initial corrosion rate, and the corrosion weight gain after 240 d is lower than that of the Zr-4 alloy. Therefore, it is more suitable for the water environment with a high concentration of dissolved oxygen. The 0Cu zirconium alloy (Comparative Example 4) is not suitable for the oxygen-rich water environment because it has two corrosion turning points, and the corrosion rate is high after the turning point, and the final weight gain is comparable to that of Nb-containing alloys such as M5 and ZIRLO.

[0088] Test Example 6

[0089] Figure 6 The corrosion weight gain curves of the zirconium alloys prepared in Examples 4-5 and Comparative Example 5 and the reference commercial alloy in the high-temperature water environment with 1000 ppb dissolved oxygen (360 °C, 20 MPa). From Figure 6 It can be seen that the stress-relieved 0.07Cu zirconium alloy (Example 5) annealed at 470 °C has the best corrosion resistance, with a low initial corrosion rate, and the weight gain after 240 d is lower than that of the Zr-4 alloy. The initial corrosion rate of the partially recrystallized 0.07Cu zirconium alloy annealed at 550 °C is relatively high, so the final weight gain (240 d) is slightly higher than that of the Zr-4 alloy. The corrosion weight gain law of the partially recrystallized 0.07Cu alloy annealed at 510 °C is similar to that of the alloy annealed at 550 °C, but its corrosion weight gain is higher than that of the alloy annealed at 550 °C. It can be seen that both the stress-relieved annealed and partially recrystallized annealed 0.07Cu zirconium alloys have good corrosion resistance and are suitable for the oxygen-rich water environment. It should be noted that in the high-concentration dissolved oxygen environment of 1000 ppb, compared with the partially recrystallized annealing, the stress-relieved annealing at a lower temperature may be more conducive to improving the corrosion resistance, which may be related to the higher matrix strength, the distribution form of SPPs, and the grain size and orientation.

[0090] Test Example 7

[0091] Table 2 Corrosion weight gain results (mg / dm 2 ) of the zirconium alloys and reference commercial alloys involved in different examples and comparative examples in different oxygen-containing water environments

[0092] 6 days 30 days 60 days 90 days 120 days 150 days 180 days 210 days 240 days Example 1 9.01 17.52 18.92 29.14 35.87 41.78 46.09 50.94 56.04 Example 2 9.98 17.22 18.22 23.13 25.61 30.75 33.96 39.54 44.63 Example 3 20.15 25.26 34.86 38.77 41.76 49.53 51.13 60.57 67.83 Example 4 22.99 28.92 39.87 40.30 40.90 47.68 50.21 59.11 64.54 Example 5 15.69 19.80 28.16 31.79 32.82 36.60 40.10 46.92 57.61 Comparative Example 1 17.02 31.76 32.85 40.79 47.05 58.07 64.59 73.19 78.80 Comparative Example 2 8.54 13.82 14.87 28.71 39.79 49.79 56.68 64.95 73.33 Comparative Example 3 10.05 14.60 26.44 28.09 29.61 32.99 35.77 41.63 47.85 Comparative Example 4 18.50 23.96 32.33 37.61 47.46 59.50 60.88 70.70 82.06 Comparative Example 5 24.43 28.92 39.87 40.30 40.90 47.68 50.21 59.11 64.54 <![CDATA[Zr-4 300ppb > 11.87 19.52 34.33 39.32 40.68 40.87 43.29 45.27 58.59 <![CDATA[M5 300ppb > 12.18 20.54 33.74 44.39 54.37 62.97 71.82 81.77 83.45 <![CDATA[ZIRLO 300ppb > 12.76 22.49 33.24 41.63 45.47 48.50 54.58 62.58 68.87 <![CDATA[Zr-4 1000ppb > 11.44 21.24 34.64 33.93 36.32 40.18 42.95 48.58 59.67 <![CDATA[M5 1000ppb > 11.50 21.48 38.52 47.09 55.33 62.34 71.01 79.86 82.46 <![CDATA[ZIRLO 1000ppb > 12.08 28.84 46.87 51.00 56.15 62.28 68.13 76.40 82.62

[0093] As can be seen from Table 2, the zirconium alloy provided by the present invention has excellent corrosion resistance in an oxygen-rich water environment. This is consistent with the results of Test Examples 4-6 and Figures 4 to 6 are the same. It should be noted that in Table 2, the oxygen-containing water environments of Examples 1-5 and Comparative Examples 1-5 are the same as those in Table 1.

[0094] Test Example 8

[0095] Figure 7 are the scanning electron microscope (SEM) cross-sectional fracture diagrams (300 ppb) of the oxide films of 0.07Cu zirconium alloy prepared for Examples 2, 4, 5 and Comparative Example 5 after 240 days of corrosion in high-temperature and high-pressure water, and the energy spectrum (EDS) chemical composition surface distribution diagrams and crystal structure analysis diagrams (1000 ppb) of the oxide films of the transmission electron microscope (TEM). Figure 7 In (a), it is the macroscopic cross-sectional morphology diagram of the oxide film of 0.07Cu zirconium alloy, and (a1) is the enlarged diagram of the area in the rectangular frame in (a), showing transverse cracks or "lamination" of the oxide film; (b) is the TEM morphology diagram of 0.07Cu zirconium alloy and the EDS surface scanning diagrams of Zr, O, Fe, Cr and Cu; (c) is the cross-sectional TEM morphology diagram of the oxide film of 0.07Cu zirconium alloy, showing the oxide film / matrix (O / M) interface and the transition layer (ZrO); (c1), (c2) and (c3) are the high-resolution TEM diagrams of the ZrO, ZrO / α-Zr interface and ZrO / ZrO2 interfaces in (c). From Figure 7 In (a), it can be seen that the oxide film of the zirconium alloy is integrally combined with the matrix, but there is a certain degree of "lamination" in the cross-sectional fracture, separated by transverse cracks, and this morphology confirms the existence of large internal stresses in the oxide film. Figure 7 In (b), the element distribution of the cross-section of the oxide film is given. The compositional contrast difference between Zr and O distinguishes the oxide film and the α-Zr matrix. The (O / M) interface between the two fluctuates, and the oxide film above some peaks generates transverse cracks due to tensile stress. Both this interface fluctuation and the transverse cracks are the results of the action of internal stresses in the oxide film. The larger enrichment areas of Fe and Cu elements respectively represent the appearance of Zr3Fe and Zr2Cu phases, which only appear in the α-Zr matrix region, indicating that these two second phases may be rapidly oxidized and dissolved into the surrounding oxide film during the corrosion process. The small enrichment areas of Fe and Cr elements represent the appearance of Zr(FeCr)2 phase, which is diffusely distributed in the α-Zr matrix and oxide film regions. This second phase delays oxidation during the corrosion process and remains in the ceramic oxide film as "particles" with metallic properties, thereby relieving the internal stress of the oxide film. It should be noted that in the oxide film region far from the O / M interface, the enrichment area of Cr is significantly more than that of Fe, indicating that when the Zr(FeCr)2 phase is finally oxidized, the Fe element diffuses into the surrounding matrix, realizing Fe 2+ or Fe 3+The "doping" of the ionic pair ZrO2 crystal inhibits the generation of oxygen vacancies, thereby reducing the diffusion of corrosive media. In addition, when these fine and dispersed Zr(FeCr)2 phases are oxidized, the stress change introduced by the local volume expansion of the oxide film is small, making the stress more "uniformly" distributed and reducing the generation of microvoids or microcracks, thereby improving the compactness and protection of the oxide film. Figure 7 (c) of the above shows the hexagonal h-ZrO existing near the O / M interface, which is the transition layer between the α-Zr matrix and the oxide film. It not only relieves the phase change stress during oxidation but also reduces the rate of the corrosion reaction. When there is a certain amount of dissolved oxygen in the corrosive medium or the α-Zr matrix itself contains a trace amount of O element, it may affect the morphological characteristics of the interface transition layer and thus affect the corrosion reaction process. Based on the above analysis, the microstructural characteristics jointly determined by the chemical composition and preparation process, such as fine and dispersed Zr(FeCr)2 phases, relatively large grain size or small orientation difference, and the O / M interface transition layer, can relieve the internal stress of the oxide film, improve the ability of the oxide film to hinder the diffusion of corrosive media, and thus improve the corrosion resistance.

[0096] The above are only the preferred embodiments of the present invention. It should be noted that for those of ordinary skill in the art, without departing from the principle of the present invention, several improvements and refinements can be made, and these improvements and refinements should also be regarded as the protection scope of the present invention.

Claims

1. A zirconium alloy for nuclear reactors, characterized in that, By weight percentage, the chemical composition is specifically: Sn 0.40 - 0.65%, Nb 0.12 - 0.25%, Fe 0.35 - 0.50%, Cr 0.15 - 0.20%, Cu 0.05 - 0.07%, O 0.08 - 0.16%, and the balance is Zr; The zirconium alloy includes an α-Zr matrix and second-phase precipitation particles; the α-Zr matrix has a hexagonal close-packed structure; the second-phase precipitation particles include Zr(FeCr)2 phase, o-Zr3Fe phase, and t-Zr2Cu phase; The average grain size of the α-Zr matrix is between 1.02 ± 1.12 and 7.57 ± 3.47 μm, and it is a stress-relieved or partially recrystallized microstructure, with the recrystallization fraction between 13% and 78%; the second-phase precipitation particles are spherical or near-spherical, and are distributed dispersively or strip-like in the grains and grain boundaries of the α-Zr matrix, with an average size between 25 and 420 nm; The preparation method of the zirconium alloy for nuclear reactors is specifically the following steps: Melting metal raw materials to obtain a zirconium alloy ingot; the composition of the zirconium alloy ingot is consistent with the chemical composition of the zirconium alloy for nuclear reactors; Hot pressing or forging the zirconium alloy ingot to obtain a blank; Successively subjecting the blank to homogenization treatment, hot rolling, cold rolling, and annealing treatment to obtain a zirconium alloy for nuclear reactors; The temperature of the annealing treatment is 470 - 550 °C, and the holding time is 3 - 7 h; the atmosphere of the annealing treatment is vacuum.

2. A method for preparing the zirconium alloy for nuclear reactors according to claim 1, specifically the following steps: Melting metal raw materials to obtain a zirconium alloy ingot; the composition of the zirconium alloy ingot is consistent with the chemical composition of the zirconium alloy for nuclear reactors according to claim 1; Hot pressing or forging the zirconium alloy ingot to obtain a blank; Successively subjecting the blank to homogenization treatment, hot rolling, cold rolling and annealing treatment to obtain a zirconium alloy for nuclear reactors; The temperature of the annealing treatment is 470 - 550 °C, and the holding time is 3 - 7 h; the atmosphere of the annealing treatment is vacuum.

3. According to the preparation method described in claim 2, characterized in that, The temperature of the hot pressing is 660 - 720 °C; the temperature of the forging is 920 - 1120 °C.

4. According to the preparation method described in claim 2, characterized in that, The temperature of the homogenization treatment is 1000 - 1070 °C, and the holding time is 0.5 - 4 h, followed by quenching to room temperature.

5. According to the preparation method described in claim 2, characterized in that, The temperature of the hot rolling is 620 - 750 °C.

6. According to the preparation method described in claim 2, characterized in that, The cold rolling is multi-pass rolling, and the reduction per pass of the cold rolling is 18 - 45%, and the intermediate annealing temperature between passes is 480 - 600 °C, and the holding time is 1 - 6 h.

7. Application of the zirconium alloy for nuclear reactors according to claim 1 or the zirconium alloy for nuclear reactors prepared by the preparation method according to any one of claims 2 - 6 in a small and micro water-cooled nuclear reactor.

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