A high-sn low-nb zirconium alloy, a preparation method and application thereof

Abstract

The application provides a high-Sn low-Nb zirconium alloy and a preparation method and application thereof, and relates to the technical field of nuclear power materials.The high-Sn low-Nb zirconium alloy provided by the application has the following chemical components in percentage by weight: Sn 1.00-1.35%, Fe 0.22-0.48%, Cr 0.12-0.20%, Nb 0-0.25% and O 0.08-0.16%, and the balance is Zr.The high-Sn low-Nb zirconium alloy provided by the application has the same or better corrosion resistance in an oxygen-rich water environment as Zircaloy alloy, and has better hydrogen absorption performance, and is suitable for an oxygen-rich high-temperature water environment; meanwhile, the high-Sn low-Nb zirconium alloy has good mechanical properties.

Description

TECHNICAL FIELD

[0001] The present application relates to the technical field of nuclear power materials, and particularly relates to a high-Sn low-Nb zirconium alloy and a preparation method and application thereof. BACKGROUND

[0002] The corrosion resistance and post-service residual plasticity of the water side of the nuclear fuel cladding are directly related to the economy, safety and advancement of the reactor. Unlike large commercial nuclear reactors, in order to simplify the system and save space, some reactor types do not use hydrogen addition and oxygen removal devices, resulting in an increase in the dissolved oxygen (DO) content in the primary loop water, and such an oxygen-rich water chemical environment will inevitably affect the corrosion resistance of the fuel cladding material. Therefore, the development of a new zirconium alloy suitable for a DO water environment is of great significance for the large-scale commercial use of small nuclear reactors, the improvement of their economy and the guarantee of their safety.

[0003] Unlike Zircaloy alloys (such as Zr-4 and Zr-2) with Sn as the main additive element, the corrosion of most Zr-Nb series (such as M5) or Zr-Sn-Nb series (such as ZIRLO) alloys in a DO water environment or high-temperature steam is significantly intensified, the corrosion rate is equivalent to that of a water loop with nucleate boiling, the oxide film may appear white "nodules", and such "nodule corrosion" can occur both in the reactor and outside the reactor. Therefore, the Nb-containing alloy is very sensitive to DO in the environment. Early oxidation studies of pure niobium showed that in an oxygen-free low-pressure steam, pure Nb follows a normal parabolic oxide film growth curve, does not undergo a corrosion transition, and the oxide film is NbO / NbO2; when a small amount of oxygen is added to the steam, pure Nb forms a porous white oxide film at an oxidation rate close to that in pure oxygen, and immediately undergoes a corrosion transition and peeling, at which time the oxide film is Nb2O5. When the corrosion weight gain is equal to the weight of the added oxygen, the oxidation rate returns to that in pure steam. This shows that: oxygenation → oxidation acceleration → oxygen depletion → oxidation rate recovery. Therefore, the protective NbO / NbO2 can form a porous Nb2O5 in a DO environment, which is due to the difference in PB ratio of the two types of oxides, and the latter produces greater volume expansion, which in turn leads to the destruction of the oxide film, thus causing accelerated corrosion and premature corrosion transition. Zr-2.5Nb alloy has the same oxidation behavior as pure Nb metal in oxygen-containing low-pressure steam, indicating that the NbO / NbO2 crystallites are converted to Nb2O5 under DO conditions, which causes sufficient stress to destroy the protection of the oxide film. The M5 and E110 alloys of 1Nb contain fine β-Nb SPPs, and the oxidation of the β-Nb SPPs to Nb2O5 in a DO water environment also causes local destruction of the oxide film, which in turn worsens the corrosion performance. Therefore, it is currently believed that Nb-containing alloys are not suitable for application in an oxygen-rich high-temperature water environment.

[0004] However, the above-mentioned DO accelerated corrosion of Nb-containing alloys is based on the oxidation of pure Nb or β-Nb SPPs into Nb2O5, and does not involve the oxidation of solid-solute Nb. Japan developed a new stress-relief annealing NDA alloy (New Developed corrosion resistant Alloy) containing 0.1% Nb, reduced Sn content (1%), and increased Fe+Cr content (0.4%), and was first successfully used in a commercial reactor refueling in 2004. The research team pointed out that the addition of 0.1-0.2% Nb made the Zr-(0.5-1.5)Sn-xNb-0.3Fe-0.1Cr alloy have a lower corrosion rate, and also confirmed the role of solid-solute Nb. In fact, the effect of Nb on the corrosion performance of zirconium alloys has two aspects: one is the delayed oxidation of β-Nb SPPs in the oxide film, and the other is the doping of different valence states of Nb (such as Nb 2+ , Nb 4+ , Nb 5+ ) into ZrO2, the former mainly causes mechanical damage due to volume expansion, and the latter mainly changes the defect type in ZrO2 and affects the conductance of corrosive media, which is an electrochemical damage. When the Nb content is controlled below its saturation solid solubility, the local mechanical damage of β-Nb SPPs oxidation into Nb2O5 will be eliminated, and only the potential electrochemical damage of Nb 5+ exists. Nb 5+ , as an n-type dopant, often exists in the outer oxide film and monoclinic ZrO2, and introduces additional oxygen vacancies, increases the migration channel of oxidizing media, and increases the corrosion rate. However, the adverse effects of Nb 5+ caused by the oxidation of solid-solute Nb may have the following two limitations: 1. The solid solubility of Nb in α-Zr is relatively low (about 0.3 at.%, 288°C), and the amount of Nb 5+ and oxygen vacancies generated after oxidation is very limited, which is negligible compared to the oxygen vacancies generated by non-stoichiometric ZrO 2-x (especially the oxide film / substrate (O / M) interface), so its role in promoting corrosion acceleration is very limited; 2. The oxidation of solid-solute Nb may be a gradual process, i.e. first oxidized into Nb 2+ or Nb 4+ , and finally oxidized into Nb 5+ after a long time of corrosion, and Nb 2+ or Nb 4+ not only inhibits oxygen transport, but also reduces the hydrogen absorption fraction. That is, a small amount of Nb can slow down corrosion, especially reduce the corrosion rate in the early corrosion stage, while also help to improve the hydrogen absorption performance.

[0005] In summary, the mechanism of DO accelerating corrosion of zirconium alloys containing Nb (in precipitated or solid solution state) is still unclear, and it seems not desirable to completely abandon the addition of Nb in the design of fuel cladding materials for DO water environment, and it may not maximize the beneficial effects of Nb (such as inhibition of hydrogen absorption). SUMMARY

[0006] The present application aims to provide a high-Sn low-Nb zirconium alloy, a preparation method and application thereof, the high-Sn low-Nb zirconium alloy provided by the present application has corrosion resistance in oxygen-rich water environment comparable to or better than Zircaloy alloy, and is suitable for oxygen-rich high-temperature water environment; meanwhile, the high-Sn low-Nb zirconium alloy has good mechanical properties.

[0007] In order to achieve the above-mentioned application purposes, the present application provides the following technical solutions:

[0008] The present application provides a high-Sn low-Nb zirconium alloy, the chemical composition of which comprises, by weight percentage: Sn 1.00-1.35%, Fe 0.22-0.48%, Cr 0.12-0.20%, Nb 0-0.25%, and O 0.08-0.16%, and the balance being Zr.

[0009] Preferably, the mass ratio of Fe and Cr (Fe / Cr) in the high-Sn low-Nb zirconium alloy is 1.0-3.0.

[0010] Preferably, the high-Sn low-Nb zirconium alloy comprises an α-Zr matrix and second phase precipitates (SPPs); the α-Zr matrix is a hexagonal close-packed structure; and the SPPs comprise Zr(Fe,Cr)2 phase.

[0011] Preferably, the SPPs are in the shape of near-ellipsoid or irregular shape, and have an average size of 25-31 nm.

[0012] Preferably, the high-Sn low-Nb zirconium alloy has a yield strength of 410±12-445±13 MPa, a tensile strength of 551±8-563±9 MPa, and an elongation of 20±1-22±2%.

[0013] The present application provides a preparation method of the high-Sn low-Nb zirconium alloy described in the above technical solutions, comprising the following steps:

[0014] The metal raw materials are subjected to melting to obtain a zirconium alloy ingot; the composition of the zirconium alloy ingot is consistent with the chemical composition of the high-Sn low-Nb zirconium alloy described in the above technical solutions.

[0015] The zirconium alloy ingot is subjected to hot pressing or forging to obtain a blank.

[0016] The blank is sequentially subjected to homogenization treatment, hot rolling, solid solution treatment, cold rolling and annealing treatment to obtain the high-Sn low-Nb zirconium alloy.

[0017] Preferably, the temperature of the solid solution treatment is 1000-1070 DEG C, the holding time is 0.5-4h, and beta phase zone quenching is performed;

[0018] The temperature of the hot rolling is 620-750 DEG C.

[0019] Preferably, the cold rolling is multi-pass rolling, the reduction of each pass is 18-45%, and vacuum annealing is performed at 480-600 DEG C between passes for 1-6h.

[0020] Preferably, the temperature of the annealing treatment is 450-600 DEG C, the holding time is 3-7h, and the atmosphere of the annealing treatment is vacuum.

[0021] The application provides application of the high-Sn low-Nb zirconium alloy in a small water-cooled nuclear reactor.

[0022] The application provides a high-Sn low-Nb zirconium alloy, which comprises, by weight percentage, Sn 1.00-1.35%, Fe 0.22-0.48%, Cr 0.12-0.20%, Nb 0-0.25% and O 0.08-0.16%, and the balance is Zr.

[0023] Sn is an alpha phase stabilizer and the only element that is almost completely in solid solution. The advantages are: Sn can increase the alpha to beta phase transition temperature, which is beneficial for hot working in the single phase alpha-Zr region; Sn can offset the detrimental effects of impurities, especially nitrogen, on corrosion resistance; Sn can increase the yield strength, tensile strength, and creep resistance through solid solution strengthening; Sn can improve the resistance to nodular corrosion in high temperature steam and dissolved oxygen (DO) environments, and avoid localized accelerated corrosion. The disadvantages are: excessive Sn is detrimental to uniform corrosion resistance in pressurized water reactor (PWR) environments, because Sn can stabilize tetragonal ZrO2 to a high volume fraction, which can cause a large local stress when the tetragonal phase transforms to monoclinic ZrO2, and thus can damage the integrity of the oxide film, causing the corrosion transition to occur prematurely; Sn can inhibit the growth of columnar grains in the oxide film (more equiaxed grains) and introduce a higher interfacial area, which can promote the diffusion of oxidizing media and increase the corrosion rate, and thus can deteriorate the corrosion resistance. Therefore, in view of the good performance of Zircaloy alloys in boiling water reactors (BWRs), and the unclear corrosion mechanism of Sn in dissolved oxygen environments (which can be different from the oxygen-free condition), and the thermal processing performance of the alloy, especially the effect of Sn on increasing the alpha to beta phase transition temperature and improving the mechanical properties of the alloy, the present application limits the Sn content to 1.00-1.35 wt. %.

[0024] Nb is a beta phase stabilizer that expands the beta phase region and promotes the formation of Nb-rich beta-Zr and stabilizes it to room temperature. As explained in the background section, Nb can be solid-solved in the alpha-Zr matrix, or can form beta-Nb second phase particles (SPPs) that precipitate. In a conventional pressurized water reactor environment, Nb can improve the corrosion resistance of the alloy, but in oxygen-containing water or high temperature steam environments, it can deteriorate the corrosion resistance. Therefore, the Nb content is selected to be below the alpha-Zr solid solution content, to avoid the formation of beta-Nb SPPs and the adverse effects of delayed oxidation. In addition, Nb can improve the irradiation growth resistance of Zr alloys and reduce the hydrogen absorption amount, so the addition of Nb is not completely abandoned, but an optimal Nb content range is sought.

[0025] Fe and Cr: Fe and Cr added in Zr alloys can be completely solid-solved in β-Zr, but the maximum solid solubility of Fe and Cr in α-phase region is only 120 ppm and 200 ppm, respectively, so they usually exist in the form of Zr(Fe,Cr)2 SPPs (Laves structure) in α-Zr. Zr(Cr,Fe)2 SPPs have two structures, hexagonal close-packed (hcp, C14) and face-centered cubic (fcc, C15), and contain typical stacking faults, and the actual crystal structure usually depends on the Fe / Cr ratio, <0.1 or >0.9 for fcc, and the intermediate ratio for hcp. The size distribution of Zr(Fe,Cr)2 SPPs is very important for the performance of Zr alloys, especially the corrosion resistance: larger SPPs size is beneficial to improve the uniform corrosion resistance in PWR environment, and small SPPs is beneficial to improve the pimple corrosion resistance in BWR. In addition, Zr(Cr,Fe)2 SPPs can also improve the performance of irradiation growth resistance and creep resistance. Considering that the Nb content in the present application is below the solid solubility of α-Zr, β-Nb SPPs will not be formed, and only Zr(Fe,Cr)2 second phase exists. Therefore, a certain amount of Fe and Cr is needed to ensure the amount of Zr(Cr,Fe)2 SPPs to ensure the mechanical properties, but larger SPPs will promote pimple corrosion, so the addition amount of Fe and Cr also needs to be controlled. In addition, reducing the Fe / Cr ratio can reduce the corrosion rate, but it increases the hydrogen absorption fraction, so the Fe / Cr ratio needs to be balanced. In view of this, the present application takes Fe 0.22-0.48% and Cr 0.12-0.20%, and controls the Fe / Cr ratio to be between 1 and 3.

[0026] O is an important alloying element in Zr alloys, usually added in the form of ZrO2 powder, with a content of 0.08-0.16wt.%. O is an α-phase stabilizing element, which can expand the α-phase region, thereby increasing the α→β phase transition temperature, and even stabilizing the α phase up to the liquidus temperature, which is beneficial to hot working in the single-phase α-Zr region at a higher temperature, but is not conducive to cold rolling deformation of Zr alloys; O mainly produces strengthening through interstitial solid solution, increasing the yield strength, and 0.1wt.% O can increase the room temperature yield strength by about 150 MPa, and the strengthening effect is not obvious at high temperature (>250℃), but can improve the irradiation growth resistance and creep performance; when the O content is high, a wider O diffusion layer may be formed at the oxide film / matrix (O / M) interface, and a ZrO transition layer is formed, which slows down the diffusion of oxidizing medium to the Zr matrix, thereby improving the corrosion resistance. In summary, in order to improve the corrosion resistance in DO environment and improve the hot working performance, the O content is preferably 0.10-0.15wt.%.

[0027] The application further provides a preparation method of the high-Sn low-Nb zirconium alloy.

[0028] Compared with the existing commercial Zr-4 alloy, the high-Sn low-Nb zirconium alloy provided by the application has excellent corrosion resistance in an oxygen-containing water environment, and is suitable for being used as a candidate material for a small-sized nuclear reactor fuel cladding, a grid strip and a structural member in an oxygen-rich water environment. BRIEF DESCRIPTION OF DRAWINGS

[0029] Figure 1 Microstructure images of zirconium alloy substrates for different embodiments and comparative examples;

[0030] Figure 2 TEM images and corresponding size distribution images of second phase precipitates (SPPs) in zirconium alloy substrates for different embodiments and comparative examples;

[0031] Figure 3 SEM morphology images of oxide film cross sections of zirconium alloys for different embodiments and comparative examples after corrosion for 150 d and 240 d in 1000 ppb dissolved oxygen (DO);

[0032] Figure 4 TEM bright field (BF) and high-angle annular dark field (HAADF) morphology images of oxide film cross sections of 0Nb and 0.6Nb zirconium alloys after corrosion for 240 d in 300 ppb and 1000 ppb DO;

[0033] Figure 5 Typical TEM bright field (BF) and high-angle annular dark field (HAADF) morphology images of oxide film cross sections of 0.1Nb alloy after corrosion for 150 d in 300 ppb DO;

[0034] Figure 6 Corrosion weight gain curves of zirconium alloys for different embodiments and comparative examples and a reference commercial alloy in different oxygen-containing water environments;

[0035] Figure 7 Corrosion weight gain curves of zirconium alloys for different embodiments and comparative examples after corrosion for 240 d with changes in Nb content and Fe / Cr ratio. DETAILED DESCRIPTION

[0036] The application provides a high-Sn low-Nb zirconium alloy, which comprises, in percentage by weight, Sn 1.00-1.35%, Fe 0.22-0.48%, Cr 0.12-0.20%, Nb 0-0.25% and O 0.08-0.16%, and the balance being Zr.

[0037] In the application, the high-Sn low-Nb zirconium alloy comprises, in percentage by weight, Sn 1.00-1.35%, preferably 1.21-1.26%.

[0038] In the application, the high-Sn low-Nb zirconium alloy comprises, in percentage by weight, Fe 0.22-0.48%, preferably 0.22-0.44%.

[0039] In the application, the high-Sn low-Nb zirconium alloy comprises, in percentage by weight, Cr 0.12-0.20%, preferably 0.16-0.18%.

[0040] In the application, the high-Sn low-Nb zirconium alloy comprises, in percentage by weight, Nb 0-0.25%, preferably 0.05-0.15%.

[0041] In the application, the high-Sn low-Nb zirconium alloy comprises, in percentage by weight, O 0.08-0.16%, preferably 0.11-0.13%.

[0042] In the application, the mass ratio of Fe and Cr (Fe / Cr) in the high-Sn low-Nb zirconium alloy is preferably 1.0-3.0, more preferably 2.0-3.0.

[0043] In the application, the high-Sn low-Nb zirconium alloy comprises, in percentage by weight, the balance of Zr. In the specific embodiment of the application, the high-Sn low-Nb zirconium alloy further comprises inevitable impurities, and the added alloying elements can avoid the adverse effects.

[0044] In the application, the high-Sn low-Nb zirconium alloy preferably comprises an α-Zr matrix and SPPs. In the application, the α-Zr matrix is preferably a hexagonal close-packed structure (hcp); the α-Zr matrix is a recrystallized structure dominated by equiaxed grains. In the application, the SPPs preferably comprise a Zr(Fe,Cr)2 phase; the SPPs preferably further comprise Nb.

[0045] In the application, the SPPs preferably have a near-ellipsoidal or irregular shape, and the average size is preferably 25-31 nm, more preferably 27 nm. In the application, the SPPs are relatively uniformly distributed, and the size difference is large.

[0046] In the present application, the high-Sn low-Nb zirconium alloy has a yield strength of 410±12-445±13 MPa, a tensile strength of 551±8-563±9 MPa, and an elongation of 20±1-22±2% at room temperature. The mechanical properties meet the material selection requirements of zirconium alloys for nuclear reactor fuel cladding.

[0047] The present application provides a preparation method of the high-Sn low-Nb zirconium alloy described in the above technical solution, comprising the following steps:

[0048] The metal raw material is melted to obtain a zirconium alloy ingot. The composition of the zirconium alloy ingot is consistent with the chemical composition of the high-Sn low-Nb zirconium alloy described in the above technical solution.

[0049] The zirconium alloy ingot is hot-pressed or forged to obtain a blank.

[0050] The blank is sequentially subjected to homogenization treatment, hot rolling, solid solution treatment, cold rolling and annealing treatment to obtain the high-Sn low-Nb zirconium alloy.

[0051] In the present application, the metal raw material preferably comprises nuclear-grade sponge zirconium and pure metal, or nuclear-grade sponge zirconium and intermediate alloy. In the present application, the melting is preferably vacuum consumable arc melting. In the present application, the temperature of the melting is preferably 1250-1885°C, the time is preferably 30-60 minutes, and the number of overturns is 3-7. The present application preferably cools to room temperature after melting to obtain a zirconium alloy ingot.

[0052] After obtaining the zirconium alloy ingot, the present application hot-presses or forges the zirconium alloy ingot to obtain a blank. In the present application, the temperature of the hot-pressing is preferably 660-720°C, and more preferably 680-700°C. In the present application, the hot-pressing has relatively uniform composition, no casting defects, and relatively low deformation temperature and load. In the present application, the hot-pressing is preferably suitable for laboratory small ingots (≤200g).

[0053] In the present application, the temperature of the forging is preferably 920-1120°C, and more preferably 950-1000°C. In the present application, the forging is preferably suitable for industrial large and medium-sized ingots.

[0054] Preferably, after the hot press forming or forging forming, the obtained material is subjected to the sequence of descaling, pickling and water washing to obtain a blank. In the present application, the pickling solution is preferably a mixed acid; the mixed acid preferably comprises hydrofluoric acid, nitric acid solution, sulfuric acid solution and water. In the present application, 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 application, the water is preferably deionized water. In the present application, the volume ratio of the hydrofluoric acid, the nitric acid solution, the sulfuric acid solution and the water is preferably 1:3:3:3. In the present application, the water washing is preferably deionized water cleaning. In the present application, the blank has a bright surface with a metallic luster.

[0055] After obtaining the blank, the blank is subjected to the sequence of homogenization treatment, hot rolling, solid solution treatment, cold rolling and annealing treatment to obtain a high-Sn low-Nb zirconium alloy. In the present application, the homogenization treatment temperature is preferably 1000-1070℃, more preferably 1050℃; the homogenization treatment holding time is preferably 0.5-4h, more preferably 1h. Preferably, after the homogenization treatment, the present application is water quenched to room temperature.

[0056] In the present application, the hot rolling temperature is preferably 620-750℃, more preferably 680℃. In the present application, the hot rolling is preferably multi-pass hot rolling, and the inter-pass interval time is preferably 5-12min; the number of hot rolling passes is preferably 4-7 passes, more preferably 5 passes; the reduction per pass is preferably 12-55%; the total reduction of the hot rolling is preferably 70-90%. Preferably, after the hot rolling, the present application is air cooled to room temperature.

[0057] In the present application, the solid solution treatment temperature is preferably 1000-1070℃, more preferably 1030℃; the holding time is preferably 0.5-4h, more preferably 2h; and then β phase zone quenching is performed. In the present application, the cooling rate of the β phase zone quenching is preferably >30℃ / s, more preferably >100℃ / s. In the present application, the β phase zone quenching is preferably water cooling. After the β phase zone quenching in the present application, the alloying elements are fully solid-solved, and in the subsequent annealing process, Fe, Cr and other alloying elements are dispersedly precipitated in the form of SPPs, and only a small amount of them is solid-solved in the α-Zr matrix.

[0058] In the present application, the cold rolling is preferably multi-pass rolling, more preferably 4-6 passes; the reduction of each pass is preferably 18-45%, more preferably 20%; the total reduction of the cold rolling is preferably 75-95%. In the present application, the passes of the cold rolling are preferably vacuum annealed at 480-600℃ for 1-6h, more preferably at 580℃ for 2h. In the present application, the cumulative annealing parameter A value (mainly affecting the microstructure and SPPs distribution characteristics of the final alloy) of the vacuum annealing is 10 -18 h; 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.

[0059] The present application obtains zirconium alloy plates or pipes of predetermined size and specification according to different rolling methods.

[0060] In the present application, the annealing temperature is preferably 450-600℃, more preferably 470-550℃; the holding time is preferably 3-7h, more preferably 5h. In the present application, the annealing atmosphere is preferably vacuum. The present application is preferably air-cooled to room temperature after the annealing.

[0061] In the present application, the recrystallization fraction of the α-Zr matrix in the high-Sn low-Nb zirconium alloy is preferably 68.13-78.56%. The high-Sn low-Nb zirconium alloy prepared by the present application is mainly composed of recrystallized structure of equiaxed crystals.

[0062] The present application provides the application of the high-Sn low-Nb zirconium alloy in the above technical solution or the high-Sn low-Nb zirconium alloy prepared by the preparation method in the above technical solution in a small-sized water-cooled nuclear reactor, preferably in a fuel cladding material for a small-sized water-cooled nuclear reactor. In the present application, the application environment is preferably an oxygen-rich high-temperature water environment; the dissolved oxygen content of the oxygen-rich high-temperature water environment is preferably 1000ppb or less, more preferably 300-1000ppb.

[0063] The high-Sn low-Nb zirconium alloy provided by the present application has uniform corrosion performance comparable to or better than that of Zr-4 alloy in a high-temperature high-pressure water environment with 300-1000ppb dissolved oxygen, and has no tendency to occur pimple corrosion after corrosion for 240d, and is suitable for use in an oxygen-rich water quality environment of a small-sized water-cooled nuclear reactor.

[0064] The technical solutions in the present application will be described clearly and completely below in combination with the embodiments in the present application. Obviously, the described embodiments are only some of the embodiments of the present application, rather than all the embodiments. Based on the embodiments in the present application, all the other embodiments obtained by a person of ordinary skill in the art without creative effort belong to the protection scope of the present application.

[0065] Examples and comparative examples

[0066] First step: alloy smelting

[0067] According to the chemical composition of the alloys in Table 1, the nuclear-grade sponge zirconium, Sn pure metal, Zr-Fe, Zr-Cr and Zr-Nb intermediate alloy raw materials are smelted into zirconium alloy ingots with a weight of 200g by using a vacuum consumable arc furnace, and then hot-pressed into a disc shape at 680℃, the surface oxide skin is removed, and then acid washing and water washing are sequentially performed to obtain a surface-brightened blank.

[0068] Second step: rolling and heat treatment

[0069] The above blank is heated to 1050℃ for 1h for homogenization treatment, and then water quenched to room temperature; then 5-pass hot rolling is performed at 680℃, the pass interval is 5min, and air cooling is performed to room temperature; then the blank is heated to 1030℃ for 0.5h for β phase zone quenching, the cooling rate is >30℃ / s (specifically, water cooling), and then 6-pass cold rolling+intermediate annealing treatment is performed, the cold rolling reduction is 20% each time, the pass interval is vacuum annealing at 580℃ for 2h, and the cumulative annealing parameter A value is 1.02x10 -18 h, and finally a 0.6mm-thick zirconium alloy plate is obtained, and a forming alloy is obtained.

[0070] Third step: final annealing

[0071] The forming alloy is heated to the final annealing temperature shown in Table 1, vacuum annealing is performed for 5h, air cooling is performed to room temperature, and a zirconium alloy is obtained.

[0072] Table 1 Alloy composition, annealing temperature and dissolved oxygen concentration in corrosion environment of examples and comparative examples

[0073]

[0074] Test example 1

[0075] Figure 1 The microstructure diagrams of the zirconium alloy (0Nb, 0.05Nb, 0.1Nb, 0.3Nb and 0.6Nb) matrices used in different examples and comparative examples are as follows: Figure 1 (a) is the EBSD grain (≥15 o grain boundary differentiation) morphology and recrystallization diagram, Figure 1(b) is the corresponding grain size distribution diagram for EBSD. Figure 1 (c) is a TEM image of the grain morphology. Figure 1 (d) is the corresponding grain size distribution diagram for TEM. It can be seen that alloys with different Nb contents all exhibit fully recrystallized grain morphology after annealing at 550℃. The rolled or quenched lath grains essentially disappear, and the structure is entirely composed of equiaxed grains. The grain sizes obtained by both analytical methods are relatively dispersed, i.e., smaller grains are distributed among large grains, forming a normal distribution. Figure 1 (b) and (d)). The EBSD analysis area is relatively large, and grain boundaries and grains are determined by differences in grain orientation. The size ranges from 2.96±1.68 to 3.58±4.78 μm, with an average of about 3 μm, while the recrystallization fraction (RF) ranges from 68.13% to 78.56%. The TEM field of view is smaller, but the grain size varies from... Figure 1 The grain sizes obtained from five similar fields of view in (c) ranged from 1.44±0.05 to 1.82±0.09 μm, with an average of about 1.5 μm. The two analytical methods yielded different grain sizes due to differences in actual resolution, with the EBSD method exhibiting a larger error. Furthermore, the Nb content had no significant effect on the grain size of the alloy matrix, showing no corresponding statistical regularity. Differences in other aggregate elements such as Sn, Fe, and Cr could all affect the fluctuations in grain size. In conclusion, alloys with different Nb contents all exhibited equiaxed recrystallized grains, with little difference in grain size and recrystallization fraction; trace amounts of Nb could not cause significant differences in grain morphology and size.

[0076] Test Example 2

[0077] Figure 2 TEM images and corresponding size distribution diagrams of second-phase precipitates (SPPs) in the zirconium alloys (0Nb, 0.05Nb, 0.1Nb, 0.3Nb, and 0.6Nb) used in different embodiments and comparative examples are shown. It can be seen that SPPs of varying sizes are distributed within the grains, mainly Zr(Fe,Cr)₂ particles, and Nb does not change their type. Due to the low Nb content and partial enrichment in Zr(Fe,Cr)₂, there is no obvious β-Nb particle precipitation. These SPPs exhibit a normal distribution with an average size between 25 and 31 nm. The Zr(Fe,Cr)₂ size gradually increases with increasing Nb content (0.1–0.6 wt.%). It is evident that Nb addition promotes Zr(Fe,Cr)₂ growth compared to the 0Nb alloy. It should be noted that the 0.05Nb alloy has larger SPP sizes than the 0.1Nb alloy, which may be attributed to its lower Fe / Cr ratio.

[0078] Test Example 3

[0079] Figure 3SEM morphologies of the cross-sections of the oxide films of the zirconium alloys (0Nb, 0.05Nb, 0.1Nb, 0.3Nb and 0.6Nb) for different examples and comparative examples corroded for 150d and 240d at 1000ppb dissolved oxygen (DO) were taken and the average thickness of the oxide film was labeled. It can be seen that the oxide film thickness of different alloys increases with the increase of corrosion time, and the corrosion is aggravated, among which the oxide film thickness of the 0.1Nb alloy is the smallest and the corrosion performance is the best. There are transverse cracks in the oxide film parallel to the oxide film / matrix (O / M) interface, especially in the oxide film corroded for 240d, indicating that the alloys have all undergone corrosion transition.

[0080] Test Example 4

[0081] Figure 4 TEM bright field (BF) and high-angle annular dark field (HAADF) morphologies of the cross-sections of the oxide films of the 0Nb and 0.6Nb zirconium alloys corroded for 240d at 300ppb and 1000ppb DO were taken and the actual weight gain was labeled. It can be seen that the typical microstructure of the oxide film of different alloys in different DO environments is composed of the outermost layer of fine and loose equiaxed crystals and relatively dense columnar crystals, and the columnar crystal region is distributed with transverse cracks of different sizes parallel to the oxide film / matrix (O / M) interface. These transverse cracks provide a rapid diffusion channel for the diffusion of the oxidizing medium, thereby accelerating the corrosion. In comparison, high DO environment accelerates the corrosion of the 0Nb alloy, but reduces the corrosion of the 0.6Nb alloy. This is because the corrosion behavior of the DO environment is determined by the Nb content and the Fe / Cr ratio, and the effects of the two factors are different at different DO contents: high Nb content is more likely to accelerate the corrosion in low DO (300ppb) environment, and high Fe / Cr is more likely to alleviate the corrosion in high DO (1000ppb) environment.

[0082] Test Example 5

[0083] Figure 5Typical TEM bright field (BF) and high angle annular dark field (HAADF) images of the cross section of the oxide film of 0.1Nb alloy corroded in 300 ppb for 150 d, and EDS area distribution maps of O, Cr, Fe and HAADF superimposed with all elements and EDS line scan curves of typical positions. It can be seen that near the oxide film / substrate (O / M) interface, the columnar crystal phase of the oxide film is relatively dense and has no obvious defects such as cracks; there is a transition layer at the O / M interface, mainly hexagonal structure zirconium monoxide h-ZrO, and the O concentration decreases from the oxide film to the substrate; the enrichment of Cr and Fe shows that there are a large number of obvious Zr(Fe,Cr)2 SPPs, and there is no obvious difference between the oxide film and the substrate, indicating that there is delayed oxidation, "inlaid" into the oxide film and plays a role in relieving stress; at the same time, EDS line scanning shows that Nb tends to diffuse to the interface between Zr(Fe,Cr)2 and the surrounding substrate, showing a "double peak" feature, but the element ratio of Cr, Fe and Nb in Zr(Fe,Cr)2 in the substrate and the oxide film does not change, indicating that no oxidation or element diffusion migration occurs.

[0084] Test Example 6

[0085] Figure 6 Corrosion weight gain curves of different embodiments and comparative examples of zirconium alloys and reference commercial alloys in different oxygen-containing water environments. Figure 7Corrosion weight gain of Zirconium alloy 240d as a function of Nb content and Fe / Cr ratio for different examples and comparative examples. From the figure, it can be seen that, (1) in 300 ppb dissolved oxygen high temperature water environment, the 0.1 Nb alloy (Example 1) has the best corrosion resistance, the initial corrosion rate is faster than the alloys with lower Nb content (0 Nb and 0.05 Nb), but the weight gain after the corrosion transition is the lowest, and the 240d weight gain is the lowest, which is comparable to the reference Zr-4 alloy. The 0.05 Nb alloy (Example 2) has the lowest initial corrosion rate, but due to the rapid increase in corrosion rate at the corrosion transition (between 60 and 120 days), the weight gain is the highest among all alloys at 120 days, and the final corrosion resistance is significantly worse than that of the 0.1 Nb alloy. The 0 Nb alloy (Example 3) has a similar weight gain rule as the 0.1 Nb alloy before the corrosion transition, but the weight gain after the transition (after 120 days) is significantly higher than that of the 0.1 Nb alloy, and then gradually slows down, and the final weight gain is only slightly higher than that of the 0.1 Nb alloy. According to the change rule of the weight gain curves of the two alloys, it can be predicted that the 0 Nb alloy will soon undergo a second transition and accelerate the weight gain, while the 0.1 Nb alloy will continue to maintain a low corrosion rate. Therefore, it can be inferred that the corrosion resistance of the 0.1 Nb alloy is better than that of the 0 Nb alloy. The corrosion rules of the above-mentioned 0 Nb, 0.05 Nb and 0.1 Nb alloys show that the corrosion weight gain does not monotonically change with the increase of Nb content, and the 0.05 Nb alloy has poor corrosion resistance. According to the composition, the 0.05 Nb alloy has the lowest Fe / Cr ratio (Table 1), which may affect the size and distribution of Zr(Fe,Cr)2SPPs, and when the SPPs are finally oxidized, the less Fe 2+ / Fe 3+ diffuses to the surrounding ZrO2 matrix, compared with the alloy with high Fe / Cr ratio, it may contain more oxygen vacancies, thereby increasing the corrosion rate. This speculation corresponds to the faster weight gain rate of the 0.05 Nb alloy at the corrosion transition, and also indirectly indicates that a large amount of low valence Fe 2+ / Fe 3+ (relative to Zr 4+) hinder the diffusion of oxidizing species in ZrO2. Therefore, it can be concluded that in the case of higher Fe / Cr ratio (>2), the addition of trace Nb (about 0.1 wt.%) is beneficial to the improvement of corrosion resistance in DO environment. The corrosion behavior of 0.3Nb alloy (Comparative Example 1) and 0.6Nb alloy (Comparative Example 2) is similar, but it seems that the 0.3Nb alloy has an earlier second corrosion transition, but the weight gain at 240 d is still lower than that of 0.6Nb, and the corrosion resistance of these two Nb alloys is inferior to that of low Nb alloy. It can be seen that, in general, the corrosion resistance in DO environment deteriorates with the increase of Nb content (>0.3 wt.%), while trace Nb (0-0.25 wt.%) is beneficial to the improvement of corrosion resistance, especially in the case of high Sn (>1 wt.%) and moderate Fe / Cr ratio (2-3). Compared with the commercial reference Zr-4 alloy, the initial corrosion rate before the corrosion transition of the Nb-containing alloys (including 0Nb) is lower than that of Zr-4, but the corrosion transition time is earlier than that of Zr-4, and the effects of the two are superimposed, and finally the weight gain at 240 d of 0Nb and 0.1Nb alloys is close to that of Zr-4, and they have good corrosion resistance in DO environment. According to the corrosion weight gain curve, Zr-4 is in the rapid weight gain stage after the first corrosion transition at 240 d, while 0Nb and 0.1Nb alloys will be in the low-speed oxidation stage after 240 d. Therefore, it can be inferred that 0Nb and 0.1Nb alloys have better corrosion resistance than Zr-4 and are suitable for application in DO environment.

[0086] (2) 1000 ppb dissolved oxygen high temperature water environment. Before the corrosion transition, the corrosion rates of 0-0.6 Nb alloys are not much different, among which the weight gain rate of 0 Nb alloy is the fastest, while that of 0.1 Nb and 0.3 Nb alloys is the lowest, but all are lower than that of Zr-4. The corrosion transition all occurs between 90-150 d, among which the weight gain at 120 d and 150 d determines the transition time and rate: at these two time points, the weight gain of 0.1 Nb alloy is the lowest, and that of 0.3 Nb alloy is the lowest, which means that the corrosion transition of 0.1 Nb alloy is the latest, while that of 0.3 Nb is the earliest, and its rate is also the fastest. After the corrosion transition, the weight gain of 0 Nb and 0.1 Nb alloys basically increases linearly, among which the weight gain of 0.1 Nb alloy at 240 d is the lowest, which is equivalent to that of Zr-4, while that of 0 Nb alloy is slightly higher; the weight gain of 0.6 Nb alloy shows a slowing trend, and the weight gain at 240 d is close to that of 0 Nb alloy; the weight gain of 0.3 Nb alloy increases slowly and then increases linearly, while that of 0.05 Nb alloy increases linearly directly, resulting in the largest weight gain at 240 d and the lowest corrosion resistance. It can be seen that only the corrosion resistance of 0.1 Nb alloy is equivalent to that of Zr-4, and the corrosion resistance of the remaining alloys is worse than that of Zr-4; the relationship between the corrosion resistance in 1000 ppb DO environment and the Nb content is less obvious than that in 300 ppb, and the corrosion resistance of 0.05 Nb alloy is the worst, while the performance of 0.6 Nb alloy is equivalent to that of 0 Nb alloy. At this time, if the effect of Fe / Cr ratio is considered, it can be found that the increase of Fe / Cr ratio improves the corrosion resistance of Nb-containing alloy (weight gain at 240 d).

[0087] Corrosion resistance of Nb-containing alloys in 300 ppb and 1000 ppb DO environments can be found that Nb content and Fe / Cr ratio have important influence: in low DO environment (300 ppb), the oxidation valence state of Nb atoms in solid solution state (Nb 5+ ) and the mass transfer effect of ZrO2 crystals determined by it play a leading role in the corrosion reaction process and corrosion rate, which shows that the corrosion resistance decreases with the increase of Nb content; while in high DO environment, Fe / Cr ratio and the distribution form of Zr(Fe,Cr)2SPPs determined by it and the diffusion and partition of Fe 2+ / Fe 3+ in the oxide film determine the corrosion reaction and reaction rate, that is, Fe 2+ / Fe 3+ can offset the effect of Nb 5+The partial adverse effect is that the corrosion resistance is improved with the increase of the Fe / Cr ratio. Finally, considering that the dissolved oxygen content is difficult to reach 1000ppb in the actual small reactor working condition, and the corrosion performance in the 300ppb environment is comprehensively considered, the current 0Nb and 0.1Nb alloy is the preferred composition of the oxygen-rich small reactor cladding material, and the 0.05Nb, 0.3Nb and 0.6Nb alloys are not suitable for the oxygen-rich water environment due to the accelerated corrosion rate after the corrosion transition. In addition, the samples with different Nb contents used in the experiment are about 0.7mm thin sheets in a fully recrystallized state, while the reference commercial Zr-4 is a partially recrystallized thin-walled tube with a wall thickness of 0.57mm (also larger in size), and the difference in sample size and heat treatment state, combined with the non-direct data of Zr-4 alloy 240d weight gain (estimated from the weight gain curve), may cause deviation in the actual corrosion weight gain. In fact, the zirconium alloy with the composition combination of 1.0-1.5wt.%Sn+0-0.7wt.%Nb+Fe / Cr (the mass ratio of Fe / Cr is 2.0-3.0) is expected to be suitable for application in the oxygen-rich water environment.

[0088] Test Example 7

[0089] Table 2 Corrosion weight gain results (mg / dm2) of the zirconium alloys involved in different examples and comparative examples and the reference commercial alloy in different oxygen-containing water environments 2 )

[0090]

[0091] As can be seen from Table 2, the zirconium alloy provided by the present application has superior corrosion resistance in the oxygen-rich water environment, and the corrosion weight gain at 240d is comparable to that of the commercial Zr-4. In the two DO environments, the corrosion performance of 0.1Nb is the best, and the final corrosion weight gain is close to that of Zr-4. In the 300ppb DO environment, the corrosion performance at 240d is: Example 1 (0.1Nb) ≈ Zr-4 > Example 3 (0Nb) > Example 2 (0.05Nb) > Comparative Example 1 (0.3Nb) > Comparative Example 2 (0.6Nb); in the 1000ppb DO environment, the corrosion performance at 240d is: Example 4 (0.1Nb) ≈ Zr-4 > Example 5 (0Nb) ≈ Comparative Example 4 (0.6Nb) > Comparative Example 3 (0.3Nb) > Comparative Example 5 (0.05Nb). It should be pointed out that in Table 2, the oxygen-containing water environment of Examples 1-5 and Comparative Examples 1-4 is consistent with Table 1.

[0092] The above only describes the preferred embodiments of the present application, and it should be pointed out that for ordinary skilled persons in the art, several improvements and refinements can be made without departing from the principles of the present application, and these improvements and refinements should also be considered as the protection scope of the present application.

Claims

1. A high-Sn, low-Nb zirconium alloy, characterized in that, The high-Sn, low-Nb zirconium alloy is composed of the following chemical components by weight percentage: Sn 1.21–1.35%, Fe 0.22–0.48%, Cr 0.12–0.20%, Nb 0.1%, and O 0.08–0.16%, with the balance being Zr; the mass ratio of Fe to Cr in the high-Sn, low-Nb zirconium alloy is 1.0–3.

0. The high-Sn, low-Nb zirconium alloy comprises an α-Zr matrix and a second phase precipitation; the α-Zr matrix has a close-packed hexagonal structure; the second phase precipitation comprises Zr(Fe,Cr)2 phase; the second phase precipitation is in a near-ellipsoidal or irregular shape with an average size of 25–31 nm. The recrystallization fraction of the α-Zr matrix in the high-Sn, low-Nb zirconium alloy is 68.13–78.56%.

2. The high-Sn, low-Nb zirconium alloy according to claim 1, characterized in that, The high-Sn, low-Nb zirconium alloy has a yield strength of 410±12~445±13MPa, a tensile strength of 551±8~563±9MPa, and an elongation of 20±1~22±2%.

3. The method for preparing the high-Sn, low-Nb zirconium alloy according to claim 1 or 2, comprising the following steps: The metal raw materials are smelted to obtain zirconium alloy ingots; the composition of the zirconium alloy ingots is consistent with the chemical composition of the high Sn low Nb zirconium alloy described in claim 1 or 2. The zirconium alloy ingot is hot-pressed or forged to obtain a billet; The billet is subjected to homogenization treatment, hot rolling, solution treatment, cold rolling and annealing treatment in sequence to obtain a high Sn and low Nb zirconium alloy.

4. The preparation method according to claim 3, characterized in that, The solution treatment temperature is 1000–1070℃, the holding time is 0.5–4h, and β-phase quenching is performed. The hot rolling temperature is 620–750°C.

5. The preparation method according to claim 3, characterized in that, The cold rolling is a multi-pass rolling process, with a reduction of 18-45% in each pass, and the cold rolling passes are vacuum annealed at 480-600℃ for 1-6 hours.

6. The preparation method according to claim 3, characterized in that, The annealing treatment is performed at a temperature of 450–600°C for 3–7 hours; the annealing atmosphere is a vacuum.

7. The application of the high Sn-low Nb zirconium alloy according to claim 1 or 2, or the high Sn-low Nb zirconium alloy prepared by the preparation method according to any one of claims 3 to 6, in a small water-cooled nuclear reactor.

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