CN116144983A - Zirconium alloy for nuclear reactor and preparation method and application thereof - Google Patents
Zirconium alloy for nuclear reactor and preparation method and application thereof Download PDFInfo
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- C22F1/00—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working
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- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22F—CHANGING THE PHYSICAL STRUCTURE OF NON-FERROUS METALS AND NON-FERROUS ALLOYS
- C22F1/00—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working
- C22F1/16—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working of other metals or alloys based thereon
- C22F1/18—High-melting or refractory metals or alloys based thereon
- C22F1/186—High-melting or refractory metals or alloys based thereon of zirconium or alloys based thereon
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- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
- G21C3/02—Fuel elements
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- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Abstract
The invention provides a zirconium alloy for a nuclear reactor, a preparation method and application thereof, and relates to the technical field of nuclear materials. The zirconium alloy provided by the invention comprises the following chemical components in percentage by weight: 0.40 to 0.65 percent of Sn, 0.12 to 0.25 percent of Nb, 0.35 to 0.50 percent of Fe, 0.15 to 0.20 percent of Cr, 0 to 0.13 percent of Cu, 0.08 to 0.16 percent of O and the balance of Zr. The invention provides a zirconium alloyFor stress relief or partial recrystallization of the structure, the α -Zr matrix is predominantly distributed with Zr (FeCr) 2 The second phase is separated out, so that the composite material not only has good mechanical properties, but also maintains excellent corrosion resistance in an oxygen-enriched high-temperature high-pressure water environment. Compared with the existing Zr alloy, the zirconium alloy provided by the invention has lower corrosion weight gain when the oxygen-enriched water environment is corroded for 240d, and meets the application of a small miniature nuclear reactor under special water chemistry conditions.
Description
Technical Field
The invention relates to the technical field of nuclear materials, in particular to a zirconium alloy for a nuclear reactor, a preparation method and application thereof.
Background
The water side corrosion resistance of the nuclear fuel cladding and the quality of residual plasticity after service are directly related to the economy, safety and advancement of the reactor. For water-cooled nuclear reactors, dissolved Oxygen (DO) in the cooling medium has a significant impact on the corrosion resistance properties of the zirconium alloy for fuel cladding. On the one hand, unlike large commercial nuclear reactors, small water-cooled reactors, in order to simplify the system or save space, some reactor designs DO not employ hydrodeoxygenation devices, resulting in an increase in DO content in the primary loop water, which tends to affect the corrosion resistance of the fuel cladding; on the other hand, after long operation, the porous scale deposited on the surface of the fuel cladding reduces the fluidity of the cooling water, and may cause an increase in the local DO concentration in addition to the enrichment of B element. Therefore, the development of the novel zirconium alloy suitable for DO water environment has important significance for large-scale commercial use of small and miniature nuclear reactors, and simultaneously improves the economy and ensures the safety of the small and miniature nuclear reactors.
At present, novel zirconium alloy is still continuously developed at home and abroad to improve the corrosion resistance, hydrogen absorption performance, mechanical performance, radiation-resistant growth performance and radiation-resistant creep performance of the zirconium alloy, wherein the corrosion resistance is the most critical one of the five performances of the zirconium alloy and is most easy to change, and the corrosion resistance of the zirconium alloy can be greatly changed by adding very trace alloy elements (ppm-level change). From the development history of zirconium alloy, alloying is a method for effectively improving the comprehensive performance of zirconium alloy, three alloy series of Zr-Sn, zr-Nb and Zr-Sn-Nb are developed based on the method, and the commercially available zirconium alloy materials of Zr-2, zr-4, E110, M5, ZIRLO, E635 and the like and novel series of zirconium alloys with application prospect represented by N36/N45 (China Nuclear industry Co., ltd.), CZ (China Nuclear Co., ltd.) and SZA (national electric investment Co., ltd.) are formed by adding alloy elements of Fe, cr, ni, cu and the like. However, these high performance zirconium alloys are designed for conventional nuclear reactor water chemistry environments where corrosion resistance properties of the zirconium alloy are generally believed to gradually decrease with increasing Nb content, contrary to the corrosion laws of conventional water chemistry environments. One possible explanation for this is that Nb oxidizes more readily to Nb under DO conditions 2 O 5 The second phase particles cause local volume expansion of the oxide film and larger additional stress, so that more crystal defects such as vacancies and the like are introduced, and the crystal defects gradually develop into micro-voids or microcracks at the columnar crystal grain boundary of the oxide film, thereby accelerating the damage to the compactness of the oxide film, promoting the interface diffusion of aggressive media and accelerating the corrosion reaction process. Another possible explanation is that Nb in a higher valence state 5+ Change ZrO 2 More oxygen vacancies are introduced to increase the erosive mediator in ZrO 2 Bulk diffusion in the crystal and migration rate in the oxide film lead to an increased corrosion rate. Therefore, these novel Nb-containing zirconium alloys may have difficulty meeting the requirements of small mini-nuclear reactors for high burnup, long cycle time and safety in oxygen-rich water environments.
For this reason, attempts have been made to develop new zirconium alloys for small and miniature water-cooled nuclear reactors based on Nb-free zr—sn based alloys such as Zr-2 or Zr-4. However, zr-2, while it has been used as a fuel cladding material for boiling water stacks, ni element therein may increase the α -Zr matrix hydrogen absorption fraction; while the Zr-4 composition does not seem to be in the optimal range, reducing Sn and increasing Fe or Fe+Cr content can greatly improve the corrosion resistance of conventional water chemistry environments. In fact, the corrosion resistance of Zr-Sn based alloys in conventional water chemistry is also significantly inferior to that of Zr-Nb or Zr-Sn-Nb based alloys. Therefore, there is still a disadvantage in that the composition is easily optimized based on the Zr-Sn alloy. Whether a small amount of Nb element can be added or whether the adverse effect of Nb element in DO water environment can be effectively restrained by compound addition of Sn, fe, cr, cu and other elements is unknown.
In summary, the existing commercial alloys and the novel zirconium alloys facing the conventional water-cooled nuclear reactor environment can not meet the requirements of the fuel cladding material of the miniature water-cooled nuclear reactor in the DO water environment to a certain extent.
Disclosure of Invention
The invention aims to provide a zirconium alloy for a nuclear reactor, a preparation method and application thereof, and the zirconium alloy provided by the invention has excellent corrosion resistance in an oxygen-enriched high-temperature high-pressure water environment, and simultaneously maintains good mechanical properties.
In order to achieve the above object, the present invention provides the following technical solutions:
the invention provides a zirconium alloy for a nuclear reactor, which comprises the following chemical components in percentage by weight: 0.40 to 0.65 percent of Sn, 0.12 to 0.25 percent of Nb, 0.35 to 0.50 percent of Fe, 0.15 to 0.20 percent of Cr, 0 to 0.13 percent of Cu, 0.08 to 0.16 percent of O and the balance of Zr.
Preferably, the zirconium alloy comprises an alpha-Zr matrix and second phase precipitate particles; the alpha-Zr matrix is of a close-packed hexagonal structure; the second phase precipitation particles include Zr (FeCr) 2 Phase, o-Zr 3 Fe phase and t-Zr 2 A Cu phase.
Preferably, the average grain size of the alpha-Zr matrix is between 1.02+/-1.12 and 7.57+/-3.47 mu m, and the recrystallization fraction is between 13 and 78 percent; the second phase precipitated particles are spherical or spheroidal and are dispersed or strip-shaped and distributed in the crystal and grain boundary of the alpha-Zr matrix, and the average size is between 25 and 420 nm.
The invention provides a preparation method of the zirconium alloy for the nuclear reactor, which comprises the following steps:
smelting a metal raw material to obtain a zirconium alloy cast ingot; the components of the zirconium alloy cast ingot are consistent with the chemical components of the zirconium alloy in the technical scheme;
performing hot press molding or forging molding on the zirconium alloy cast ingot to obtain a blank;
and sequentially carrying out homogenization treatment, hot rolling, cold rolling and annealing treatment on the blank to obtain the zirconium alloy for the nuclear reactor.
Preferably, the temperature of the hot press molding is 660-720 ℃; the forging forming temperature is 920-1120 ℃.
Preferably, the homogenization treatment is carried out at a temperature of 1000-1070 ℃ for 0.5-4 hours, and then quenching is carried out to room temperature.
Preferably, the temperature of the hot rolling is 620 to 750 ℃.
Preferably, the cold rolling is multi-pass rolling, the reduction of each pass of cold rolling is 18-45%, the intermediate annealing temperature between passes is 480-600 ℃, and the heat preservation time is 1-6 h.
Preferably, the temperature of the annealing treatment is 450-600 ℃, and the heat preservation time is 3-7 h; the atmosphere of the annealing treatment is vacuum.
The invention provides the application of the zirconium alloy for the nuclear reactor or the zirconium alloy for the nuclear reactor prepared by the preparation method in the technical scheme in the small and miniature water-cooled nuclear reactor.
The invention provides a zirconium alloy for a nuclear reactor, which comprises the following chemical components in percentage by weight: 0.40 to 0.65 percent of Sn, 0.12 to 0.25 percent of Nb, 0.35 to 0.50 percent of Fe, 0.15 to 0.20 percent of Cr, 0 to 0.13 percent of Cu, 0.08 to 0.16 percent of O and the balance of Zr. In the invention, the functions and the component selection of various additive elements are based on the following:
sn can improve the phase transition point of the zirconium alloy, is convenient for processing deformation in an alpha-Zr single-phase region at higher temperature, and reduces or avoids residual high-temperature beta-Zr phase in room temperature tissues. Meanwhile, due to the segregation of solid solution Sn at the defect positions of oxide film grain boundaries and the like, sn can counteract the harmful effect of N, C, al and other impurity elements in the sponge zirconium on corrosion resistance; thanks to solid solution strengthening in alpha-Zr, sn can improve the mechanical properties of the zirconium alloy such as yield strength, tensile strength, creep resistance and the like; in addition, sn can balance furuncle corrosion resistance and irradiation creep resistance. However, excessive Sn may increase the tetragonal structure t-ZrO during etching 2 When it is oriented toward monoclinic structure m-ZrO 2 More additional stress is caused during the transition of the oxide film, and the compactness of the oxide film is destroyed, thereby inducing corrosion turning and reducing corrosion resistance. At present, the high-performance zirconium alloy for the core is generally designed by adopting a low Sn component. Therefore, the Sn content is controlled at a lower level (lower than commercial Zr-4, ZIRLO and the like alloys).
Nb may be solid-dissolved in the α -Zr matrix or may form β -Nb second phase precipitate particles, but the maximum solid solubility of Nb in α -Zr is affected by other alloying elements. In a conventional pressurized water reactor primary circuit water environment, nb can greatly improve the corrosion resistance of the alloy, whereas in an oxygen-rich water or high temperature steam environment, nb can deteriorate the corrosion performance. For this reason, the Nb content is selected to be lower than the solid solution content in alpha-Zr, avoiding the adverse effect of beta-Nb particles due to delayed oxidation in the oxide film. In addition, since Nb can improve the performance of zirconium alloy such as radiation-resistant growth, the addition of Nb is not completely abandoned, but SPPs are generated by the addition of other alloying elements such as Fe and Cr to minimize the damage of solid solution Nb to corrosion resistance.
The addition of Fe firstly can avoid the decrease of mechanical properties of the alloy caused by the decrease of Sn and Nb contents, and secondly can promote Zr (Fe, cr) 2 The generation of the second phase can improve the hydrogen absorption performance, furuncle corrosion resistance and radiation growth resistance of the alloy matrix. However, excessive amounts of Fe may form a corrosion resistance disadvantageZr of larger energy size 3 And Fe phase.
Cr has similar effect with Fe, and can improve comprehensive mechanical properties, and simultaneously form Zr (Fe, cr) by being matched with Fe 2 Second phase, and improves furuncle corrosion resistance, radiation growth resistance and creep resistance of the alloy, especially improves corrosion resistance of the weld joint of the cladding and the end plug. However, too much Cr is detrimental to the processing of the zirconium alloy and also increases the hydrogen absorption fraction of the alloy.
O has a high solid solubility in α -Zr, up to 29at.%. Therefore, on one hand, the thickness and the shape of the h-ZrO transition layer and the O supersaturated solid solution layer of the hexagonal structure near the interface of the oxide film/matrix (O/M) in the corrosion process can be influenced, so that the oxidation reaction process is influenced; on the other hand, O can change the strength of the oxidized front-end metal matrix through solid solution strengthening, and influences the stress relaxation degree in the oxide film, thereby influencing corrosion turning. In addition, O can improve the radiation-resistant growth and creep resistance of the alloy. However, too much O is detrimental to the processing of the zirconium alloy.
Cu diffuses at a slower rate than Fe in alpha-Zr and tends to replace Zr (Fe, cr) 2 Fe in the phase and is enriched therein, thus altering the kinetics of the precipitation reaction such that Zr (Fe, cr) 2 The phase is finer and dispersed; trace Cu can also inhibit Zr 3 The Fe phase grows and coarsens excessively, avoiding it generating more additional stress during oxidation. However, excessive Cu may form coarser Zr 2 The Cu phase damages the compactness by introducing larger local stress into the oxide film, thereby deteriorating the corrosion resistance. In addition, cu reduces the recrystallization temperature of the alloy, widens the processing process window of the alloy, increases the recrystallization fraction of alpha-Zr during final annealing, and reduces the equivalent grain size thereof. In addition, cu in solid solution can inhibit Nb in high valence state 5+ And the formation of the corrosion reaction rate in the oxygen-enriched water environment is slowed down.
In addition to the single action of the above-mentioned alloying elements, coupling action of a plurality of elements can be generated in the actual alloy, and advantageous effects can be generated. For example, nb is in a solid solution state completely below the limit solid solubility, and Zr (Fe, cr) is generated when Fe, cr are present 2 During the phase, nbCan diffuse and concentrate in the water to further reduce the solid solution content; meanwhile, when Cu is present, nb and Cu both suppress Zr (Fe, cr) because the diffusion rate in alpha-Zr is lower than Fe 2 The phase grows up, and the SPPs distribution of fine dispersion is obtained. In addition, cu can reduce adverse effects even if a very small amount of solid-dissolved Nb is present in the alpha-Zr. Therefore, as long as the component interval selection is reasonable, the coupling addition of various elements can not only optimize the microstructures such as SPPs and the like, but also inhibit the adverse effects of certain elements (such as Nb in an oxygen-enriched water environment), thereby achieving the purposes of simultaneously improving the mechanical property and the corrosion resistance, and expanding the application of the zirconium alloy in special water chemistry.
The zirconium alloy provided by the invention adopts the component design of 'medium Sn low Nb' and contains proper Cu by combining the single and coupling effects of the elements, fully considering the application environment of oxygen-enriched high-temperature high-pressure water, so that the Zr-Sn-Fe-Cr alloy has excellent corrosion resistance in an oxygen-enriched water environment while maintaining good mechanical properties, and meets the requirements of a small miniature water-cooling nuclear reactor on improving economy and safety.
The results of the examples show that the zirconium alloy provided by the invention has excellent corrosion resistance in an oxygen-enriched high-temperature high-pressure water environment, and compared with the prior commercial M5, ZIRLO and Zr-4 alloys, the zirconium alloy has lower corrosion weight gain when the corrosion of two oxygen-enriched water environments of 300ppb and 1000ppb is up to 240d, and is suitable for being used as a fuel cladding, a grid strip and a structural member of a small miniature nuclear reactor.
Drawings
FIG. 1 is a graph showing the morphology and size distribution of Electron Back Scattering Diffraction (EBSD) of zirconium alloys prepared in examples 1-2 and comparative examples 1-2;
FIG. 2 is a graph of the microstructure of a Scanning Electron Microscope (SEM) and a Transmission Electron Microscope (TEM) of the zirconium alloys prepared in examples 1-2 and comparative examples 1-2;
FIG. 3 is a Transmission Electron Microscope (TEM) morphology and corresponding selected electron diffraction (SAED) pattern and energy spectrum (EDS) composition analysis of the second phase of the zirconium alloys prepared in examples 1 and 3;
FIG. 4 is a graph showing a comparison of corrosion weight gain curves of the zirconium alloys prepared in examples 1-2 and comparative examples 1-2 in a 300ppb dissolved oxygen high temperature high pressure aqueous environment (360 ℃ C., 20 MPa);
FIG. 5 is a graph comparing corrosion weight gain curves of zirconium alloys prepared in examples 3-4 and comparative examples 3-4 in a 1000ppb dissolved oxygen high temperature and pressure aqueous environment (360 ℃, 20 MPa);
FIG. 6 is a graph comparing corrosion weight gain curves of zirconium alloys prepared in examples 4-5 and comparative example 5 in a 1000ppb dissolved oxygen high temperature and pressure aqueous environment (360 ℃, 20 MPa);
FIG. 7 is a Scanning Electron Microscope (SEM) cross-sectional fracture map (300 ppb) and a Transmission Electron Microscope (TEM) oxide film energy spectrum (EDS) chemical composition plane distribution map and a crystal structure analysis map (1000 ppb) of the 0.07Cu zirconium alloy prepared in examples 2, 4, 5 and comparative example 5 corroded 240d oxide film in high temperature and high pressure water.
Detailed Description
The invention provides a zirconium alloy for a nuclear reactor, which comprises the following chemical components in percentage by weight: 0.40 to 0.65 percent of Sn, 0.12 to 0.25 percent of Nb, 0.35 to 0.50 percent of Fe, 0.15 to 0.20 percent of Cr, 0 to 0.13 percent of Cu, 0.08 to 0.16 percent of O and the balance of Zr.
The zirconium alloy for nuclear reactor provided by the invention comprises 0.40-0.65% of Sn, preferably 0.45-0.50% by weight. According to the invention, sn can improve the phase transition point of the zirconium alloy, is convenient for processing deformation in an alpha-Zr single-phase region at a higher temperature, and reduces or avoids residual high-temperature beta-Zr phase in a room-temperature structure. Meanwhile, due to the segregation of solid solution Sn at the defect positions of oxide film grain boundaries and the like, sn can counteract the harmful effect of N, C, al and other impurity elements in the sponge zirconium on corrosion resistance; thanks to solid solution strengthening in alpha-Zr, sn can improve the mechanical properties of the zirconium alloy such as yield strength, tensile strength, creep resistance and the like; in addition, sn can balance furuncle corrosion resistance and irradiation creep resistance. However, excessive Sn may increase the tetragonal structure t-ZrO during etching 2 When it is oriented toward monoclinic structure m-ZrO 2 Causes a larger forehead when the transition of (a)External stress damages the compactness of the oxide film, thereby inducing corrosion turning and reducing corrosion resistance. At present, the high-performance zirconium alloy for the core is generally designed by adopting a low Sn component. Therefore, the Sn content is controlled at a lower level (lower than commercial Zr-4, ZIRLO and the like alloys).
The zirconium alloy for nuclear reactor provided by the invention comprises 0.12-0.25% of Nb, preferably 0.15-0.20% by weight. In the present invention, nb may be solid-dissolved in the α -Zr matrix or may form β -Nb second phase precipitated particles, but the maximum solid solubility of Nb in α -Zr is affected by other alloying elements. In a conventional pressurized water reactor primary circuit water environment, nb can greatly improve the corrosion resistance of the alloy, whereas in an oxygen-rich water or high temperature steam environment, nb can deteriorate the corrosion performance. For this reason, the Nb content is selected to be lower than the solid solution content in alpha-Zr, avoiding the adverse effect of beta-Nb particles due to delayed oxidation in the oxide film. In addition, since Nb can improve the performance of zirconium alloy such as radiation-resistant growth, the addition of Nb is not completely abandoned, but SPPs are generated by the addition of other alloying elements such as Fe and Cr to minimize the damage of solid solution Nb to corrosion resistance.
The zirconium alloy for nuclear reactor provided by the invention comprises 0.35-0.50% of Fe, preferably 0.36-0.49% by weight. In the invention, the addition of Fe firstly can avoid the decrease of mechanical properties of the alloy caused by the decrease of Sn and Nb contents, and secondly can promote Zr (Fe, cr) 2 The generation of the second phase can improve the hydrogen absorption performance, furuncle corrosion resistance and radiation growth resistance of the alloy matrix. However, excessive amounts of Fe may form larger-sized Zr that is detrimental to corrosion resistance 3 And Fe phase.
The zirconium alloy for nuclear reactor provided by the invention comprises 0.15-0.20% of Cr, preferably 0.18-0.19% of Cr. In the invention, the effect of Cr and Fe is similar, the comprehensive mechanical property can be improved, and Zr (Fe, cr) is formed by matching with Fe 2 Second phase, and improves furuncle corrosion resistance, radiation growth resistance and creep resistance of the alloy, especially improves corrosion resistance of the weld joint of the cladding and the end plug. However, too much Cr is detrimental to the processing of the zirconium alloy and increasesAnd adding hydrogen absorption fraction of the alloy.
The zirconium alloy for nuclear reactor provided by the invention comprises 0-0.13% of Cu, preferably 0.02-0.10%, more preferably 0.05-0.07% by weight. In the present invention, cu diffuses at a slower rate than Fe in alpha-Zr and tends to replace Zr (Fe, cr) 2 Fe in the phase and is enriched therein, thus altering the kinetics of the precipitation reaction such that Zr (Fe, cr) 2 The phase is finer and dispersed; trace Cu can also inhibit Zr 3 The Fe phase grows and coarsens excessively, avoiding it generating more additional stress during oxidation. However, excessive Cu may form coarser Zr 2 The Cu phase damages the compactness by introducing larger local stress into the oxide film, thereby deteriorating the corrosion resistance. In addition, cu reduces the recrystallization temperature of the alloy, widens the processing process window of the alloy, increases the recrystallization fraction of alpha-Zr during final annealing, and reduces the equivalent grain size thereof. In addition, cu in solid solution can inhibit Nb in high valence state 5+ And the formation of the corrosion reaction rate in the oxygen-enriched water environment is slowed down. The invention controls the Cu content in the above range, can obtain proper alpha-Zr matrix grain size and recrystallization fraction, and obtain finer dispersed Zr (FeCr) 2 Phase, avoid larger size Zr 3 Fe phase and Zr 2 The Cu phase is generated in large quantities.
The zirconium alloy for nuclear reactor provided by the invention comprises 0.08-0.16% of O, preferably 0.12-0.13% by weight. In the present invention, O has a high solid solubility in α -Zr, up to 29at.%. Therefore, on one hand, the thickness and the shape of the h-ZrO transition layer and the O supersaturated solid solution layer of the hexagonal structure near the interface of the oxide film/matrix (O/M) in the corrosion process can be influenced, so that the oxidation reaction process is influenced; on the other hand, O can change the strength of the oxidized front-end metal matrix through solid solution strengthening, and influences the stress relaxation degree in the oxide film, thereby influencing corrosion turning. In addition, O can improve the radiation-resistant growth and creep resistance of the alloy. However, too much O is detrimental to the processing of the zirconium alloy.
The zirconium alloy for the nuclear reactor provided by the invention comprises the balance of Zr, in particular to nuclear grade zirconium sponge for removing Hf. In a specific embodiment of the present invention, the zirconium alloy further comprises unavoidable impurities, and the added alloying elements may avoid adverse effects thereof.
The invention fully utilizes the single and coupling actions of the alloy elements and has beneficial effects. For example, nb is in a solid solution state completely below the limit solid solubility, and Zr (Fe, cr) is generated when Fe, cr are present 2 During phase, nb can diffuse and concentrate in the phase, so that the solid solution content of the phase is further reduced; meanwhile, when Cu is present, nb and Cu both suppress Zr (Fe, cr) because the diffusion rate in alpha-Zr is lower than Fe 2 The phase grows up, and the SPPs distribution of fine dispersion is obtained. In addition, cu can reduce adverse effects even if a very small amount of solid-dissolved Nb is present in the alpha-Zr. Therefore, as long as the component interval selection is reasonable, the coupling addition of various elements can not only optimize the microstructures such as SPPs and the like, but also inhibit the adverse effects of certain elements (such as Nb in an oxygen-enriched water environment), thereby achieving the purposes of simultaneously improving the mechanical property and the corrosion resistance, and expanding the application of the zirconium alloy in special water chemistry.
In the present invention, the microstructure of the zirconium alloy preferably includes an a-Zr matrix and second phase precipitation particles (SPPs).
In the invention, the alpha-Zr matrix is a destressing or partially recrystallized microstructure, the recrystallization fraction is between 13 and 78 percent, and the average grain size is between 1.02+/-1.12 and 7.57+/-3.47 mu m. The α -Zr matrix is preferably in the form of closely packed hexagonal structures (hcp) and β -Zr in the form of a body-centered cubic structure (bcc) is present.
In the present invention, the second phase precipitate particles preferably include Zr (FeCr) having hcp and face-centered cubic (fcc) structures 2 Phase and possible o-Zr of bottom-centre orthogonal structure 3 t-Zr of Fe phase and tetragonal structure 2 Cu phase, wherein hcp-/fcc-Zr (FeCr) 2 The phase is mainly, the volume ratio is preferably 78-95 vol.%, the phase is spherical or spheroidal, the phase is dispersed or banded in the crystal and grain boundary of the alpha-Zr matrix, and the average size is 25-420 nm, preferably 25-120 nm.
In the present invention, cu is important for the microstructureIs effective in (1). The addition of trace Cu improves the average grain size and recrystallization fraction of the alpha-Zr matrix, and changes the distribution, type, size and composition of SPPs. SPPs are mainly o-Zr with larger size when no Cu is added 3 Fe phase and small amount of fine fcc-Zr (FeCr) 2 A phase; when trace Cu is added, cu replaces part of Fe and enriches Zr (FeCr) 2 The phase further influences the precipitation kinetics, refines the size, and has a small amount of Cu-rich o-Zr 3 Fe phase and Fe-rich t-Zr 2 A Cu phase; when Cu is excessively added, t-Zr with larger size is generated 2 The Cu phase causes an increase in the average size of the second phase. Thus, the Cu content is controlled to optimize the microstructure of the alloy.
The invention provides a preparation method of the zirconium alloy for the nuclear reactor, which comprises the following steps:
smelting a metal raw material to obtain a zirconium alloy cast ingot; the components of the zirconium alloy cast ingot are consistent with the chemical components of the zirconium alloy in the technical scheme;
performing hot press molding or forging molding on the zirconium alloy cast ingot to obtain a blank;
and sequentially carrying out homogenization treatment, hot rolling, cold rolling and annealing treatment on the blank to obtain the zirconium alloy for the nuclear reactor.
Smelting a metal raw material to obtain a zirconium alloy cast ingot; the components of the zirconium alloy cast ingot are consistent with the chemical components of the zirconium alloy in the technical scheme. In the present invention, the metal raw material preferably includes nuclear grade zirconium sponge and a pure metal, or nuclear grade zirconium sponge and a master alloy. In the present invention, the smelting is preferably vacuum consumable arc smelting. In the invention, the smelting temperature is preferably 1250-1885 ℃, the time is preferably 30-60 minutes, and the turnover times are 3-7. The invention preferably carries out air cooling to room temperature after smelting to obtain the zirconium alloy cast ingot.
After the zirconium alloy cast ingot is obtained, the zirconium alloy cast ingot is subjected to hot press molding or forging molding to obtain a blank. In the present invention, the temperature of the hot press molding is preferably 660 to 720 ℃, more preferably 680 to 700 ℃. In the invention, the hot-press molding components are relatively uniform, no casting defect exists, the deformation temperature and the load are relatively low, and the hot-press molding method is suitable for small-sized laboratory ingots (less than or equal to 200 g).
In the present invention, the forging forming temperature is preferably 920 to 1120 ℃, more preferably 950 to 1000 ℃. In the present invention, the forging forming is preferably applied to industrial-grade large and medium-sized ingots.
In the invention, the obtained material is subjected to descaling, acid washing and water washing in sequence after the hot press molding or the forging molding to obtain a blank. In the invention, the washing liquid used for the acid washing is preferably 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 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 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 deionized water washing. In the present invention, the surface of the blank is bright and has metallic luster.
After the blank is obtained, the blank is sequentially subjected to homogenization treatment, hot rolling, cold rolling and annealing treatment to obtain the zirconium alloy for the nuclear reactor. In the present invention, the temperature of the homogenization treatment is preferably 1000 to 1070 ℃, more preferably 1050 ℃, in the β -Zr single-phase region; the incubation time for the homogenization treatment is preferably 0.5 to 4 hours, more preferably 1 hour. The invention preferably after said homogenization treatment is quenched to room temperature, preferably at a cooling rate of > 30 ℃/s, more preferably > 100 ℃/s. In the present invention, the quenching is preferably performed by water cooling. After beta-phase quenching, the alloy elements are fully dissolved, and in the subsequent annealing treatment process, the alloy elements such as Fe, cr and the like are dispersed and separated out in the form of SPPs, and only a very small amount of the alloy elements are dissolved in an alpha-Zr matrix in a solid manner.
In the present invention, the temperature of the hot rolling is preferably 620 to 750 ℃, more preferably 680 ℃. In the invention, the hot rolling is preferably multi-pass hot rolling, and the pass interval time is preferably 5-12 min; the number of hot rolling passes is preferably 4 to 7 passes, more preferably 5 passes; the reduction of each pass of hot rolling is preferably 12-55%; the total reduction of the hot rolling is preferably 70 to 90%. In the present invention, it is preferable that the hot rolled sheet obtained after the hot rolling is air-cooled to room temperature.
In the present invention, the cold rolling is preferably multi-pass rolling, more preferably 4 to 6 passes; the reduction of each cold rolling pass is preferably 18-45%, more preferably 20%; the total reduction of the cold rolling is preferably 75 to 95%. In the present invention, the cold rolling is preferably vacuum annealed at 480 to 600 ℃ for 1 to 6 hours, more preferably at 580 ℃ for 2 hours. In the present invention, the cumulative annealing parameter A value of the vacuum annealing (mainly affecting the microstructure and SPPs distribution characteristics of the final alloy) is 10 -18 h, performing H; the a= Σt i ×exp(-Q/RT i ) Q/r=40000K; t is annealing temperature, and the unit is K; t is annealing time, and the unit is h; q is the activation energy.
According to different rolling modes, the zirconium alloy plate or tube with preset size and specification is obtained.
In the present invention, the temperature of the annealing treatment is preferably 450 to 600 ℃, more preferably 470 to 550 ℃; the holding time is preferably 3 to 7 hours, more preferably 5 hours. In the present invention, the atmosphere of the annealing treatment is preferably vacuum. In the present invention, it is preferable that the annealing treatment is followed by air cooling to room temperature.
In the present invention, when the annealing treatment temperature is 550 ℃, the highest recrystallization fraction of the alpha-Zr matrix is 77.2%, which is a partially recrystallized structure composed of equiaxed crystal and lath crystal grains; when the temperature of the annealing treatment was 470 ℃, the recrystallization fraction was 13%, and the α -Zr matrix was considered to be a stress-relieved annealed structure dominated by lath-like grains.
The yield strength of the partially recrystallized crystalline zirconium alloy (the annealing treatment temperature is 550 ℃ or 510 ℃) is 390+/-12-425+/-13 MPa, the tensile strength is 542+/-9-558+/-8 MPa, and the elongation is 22+/-1-24+/-2%; the stress-free zirconium alloy (the annealing treatment temperature is 470 ℃) has yield strength of 565+/-9-589+/-10 MPa, tensile strength of 772+/-6-792+/-6 MPa and elongation of 16+/-1-18+/-1%. The mechanical property meets the material selection requirement of the zirconium alloy for the nuclear reactor fuel cladding.
The invention also provides application of the zirconium alloy for the nuclear reactor or the zirconium alloy for the nuclear reactor obtained by the preparation method of the technical scheme in small miniature water-cooled nuclear reactors, preferably in fuel cladding materials, and more preferably in spacer grids or plate-shaped members for tubular fuel cladding or fuel cladding tubes. In the invention, in order to evaluate the applicability of the zirconium alloy in a small miniature water-cooled nuclear reactor, the experimental environment is preferably oxygen-enriched high-temperature high-pressure water; the dissolved oxygen content of the oxygen-enriched water is preferably 300-1000 ppb so as to simulate a severe strong oxidizing environment and accelerate the evaluation process; the temperature of the high-temperature high-pressure water environment is preferably 360 ℃, and the pressure is preferably 20MPa.
The low Nb Cu-containing zirconium alloy provided by the invention has uniform corrosion performance in 300-1000 ppb dissolved oxygen high-temperature and high-pressure water environment, is equivalent to that of Zr-2 or Zr-4 and other zirconium alloy without Nb, is superior to that of commercial Zr-Nb series M5, zr-Sn-Nb series ZIRLO and other alloys, has no tendency of furuncle corrosion during 240d corrosion, and is suitable for oxygen-enriched water environment of small miniature water-cooled nuclear reactor.
The technical solutions of the present invention will be clearly and completely described in the following in connection with the embodiments of the present invention. It will be apparent that the described embodiments are only some, but not all, embodiments of the invention. All other embodiments, which can be made by those skilled in the art based on the embodiments of the invention without making any inventive effort, are intended to be within the scope of the invention.
Examples 1 to 5 and comparative examples 1 to 5
The first step: alloy smelting
According to the chemical compositions of the alloys in Table 1, nuclear grade sponge zirconium, sn pure metal, zr-Cu, zr-Fe, zr-Cr and Zr-Nb intermediate alloy raw materials are smelted into zirconium alloy cast ingots by a vacuum consumable arc furnace, the weight is 200g, the zirconium alloy cast ingots are hot pressed into discs at 680 ℃, surface oxide skin is removed, and acid washing and water washing are sequentially carried out to obtain blanks with bright surfaces.
And a second step of: rolling and heat treatment
Heating the blank to 1050 ℃, preserving heat for 1h, homogenizing, and quenching to room temperature; then carrying out 5-pass hot rolling at 680 ℃ with 5-min pass intervals, and air cooling to room temperature; then cold rolling and intermediate annealing treatment are carried out for 6 times, the reduction of each cold rolling is 20%, vacuum annealing is carried out for 2 hours at 580 ℃ among the passes, and the cumulative annealing parameter A value is 1.02X10 -18 And h, finally obtaining the zirconium alloy plate with the thickness of 0.6mm to obtain the formed alloy.
And a third step of: final annealing
And heating the formed alloy to the final annealing temperature shown in table 1, preserving heat for 5 hours, carrying out vacuum annealing, and cooling to room temperature by air to obtain the zirconium alloy.
TABLE 1 alloy compositions, annealing temperatures and dissolved oxygen concentrations for examples 1 to 5 and comparative examples 1 to 5
Test example 1
FIG. 1 is a graph showing the morphology of Electron Back Scattering Diffraction (EBSD) and corresponding size distribution of zirconium alloy substrates of different Cu contents at 550℃for 5 hours in examples 1-2 and comparative examples 1-2. As can be seen from FIG. 1, in the absence of Cu addition (comparative example 2), there was a remarkable lath-like non-recrystallized structure in the zirconium alloy, at which the recrystallization fraction was 55.9%, and the equivalent grain size was 1.14.+ -. 1.12. Mu.m; with the addition of Cu (example 1) and the increase in content (example 2 and comparative example 1), the recrystallization fraction increased from 61.4% to 77.2%, while the grain size increased from 1.53.+ -. 1.96 μm to 7.57.+ -. 3.47. Mu.m. It can be seen that under the same preparation process and final annealing conditions, both the degree of recrystallization and the average grain size increase with increasing Cu content, i.e. the addition of Cu favors the obtaining of larger sized equiaxed grains. At the same time, there is some difference in orientation of these grains. When the zirconium alloy is corroded, the higher the grain boundary ratio of the α—zr matrix, the larger the difference in crystal orientation, and the easier the corrosion reaction is promoted. 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 the orientation difference.
Test example 2
FIG. 2 is a Scanning Electron Micrograph (SEM) and a Transmission Electron Micrograph (TEM) of second phase particle precipitates (SPPs) in zirconium alloys having different Cu contents at 550℃for 5 hours in examples 1 to 2 and comparative examples 1 to 2. From the aspect of SEM morphology, the SPPs of the 0Cu and 0.05Cu zirconium alloys are similar in size distribution, fine and dispersed, while the SPPs in the 0.07Cu and 0.13Cu zirconium alloys are obviously increased in size, and the number density is obviously reduced, and are mainly distributed at grain boundaries. From the TEM morphology, the grain morphology distribution was consistent with the EBSD results in test example 1, i.e., the 0Cu zirconium alloy had the lowest recrystallization degree, and a large number of defects such as dislocations were distributed, whereas as the Cu content increased, the grain size increased and defects such as dislocations decreased. From the TEM morphology, the 0Cu zirconium alloy has more SPPs, irregular distribution and larger size difference, and different types of SPPs can exist; when 0.05wt.% Cu is added, precipitation of large-sized SPPs is suppressed, the number is significantly reduced, and finer SPPs appear only in the crystal; as the Cu content increases to 0.07wt.% and 0.13wt.%, fine dispersed SPPs are formed within the crystal, and SPPs at the individual grain boundaries are relatively large. Therefore, trace Cu can inhibit coarsening of SPPs in Zr-Sn-Nb-Fe-Cr alloy, improve distribution form of SPPs and enable the SPPs to be more finely dispersed (relative to Cu-free zirconium alloy). The size and distribution differences of such SPPs may affect their oxidation-retarding properties in the oxide film: on the one hand, the stress concentration is reduced by improving the stress distribution in the oxide film through deformation while maintaining the metal property, and on the other hand, the additional stress is applied to the surrounding oxide film due to the higher proportion of the rolling-Bedworth (P-B) to the volume expansion during final oxidation, thereby influencing the defects and the integrity in the oxide film. In a word, the addition of a proper amount of Cu promotes the SPPs to be more finely and dispersedly separated out, so that the corrosion resistance of the zirconium alloy is improved.
Test example 3
FIG. 3 is a Transmission Electron Microscope (TEM) morphology and corresponding selected electron diffraction (SAED) pattern and energy spectrum (EDS) composition analysis of the second phase of the zirconium alloys prepared in examples 1 and 3. FIG. 3 (a) shows a TEM bright field morphology of the precipitated phase SPP1 and EDS components (a) 1 ) Is thatSAED patterns corresponding to SPP1, calibrated as
(b) TEM bright field phase morphology and EDS component (b) for precipitated phase SPP2 1 ) SAED patterns for corresponding SPP2, calibrated as
(c) TEM bright field phase morphology and EDS component (c) for precipitated phase SPP3 1 ) SAED pattern for the corresponding SPP3, scaled as +.>
(d) TEM bright field phase morphology and EDS component (d) for precipitated phase SPP4 1 ) SAED pattern for the corresponding SPP4, scaled as +.>
(e) TEM bright field phase morphology and EDS component (e) for precipitated phase SPP5 1 ) SAED pattern for the corresponding SPP5, scaled as +.>
(f) TEM bright field phase morphology and EDS component for precipitated phase SPP6, (f) 1 ) SAED patterns for the corresponding SPP6, calibrated as
As can be seen, in the alloy containing 0.05Cu, there are three types of second phase precipitates, zr (FeCr), respectively 2 Phase, cu-rich Zr 3 Fe phase and Fe-rich Zr 2 A Cu phase. SAED has shown that Zr (FeCr) 2 The phases have two structures, namely a close-packed hexagonal (hcp) structure and a face centered cubic (fcc) structure, zr 3 Fe phase is a bottom center orthogonal structure (o), zr 2 The Cu phase is of tetragonal structure (t). EDS analysis showed Zr (FeCr) 2 The phase contains a certain amount of Nb, while Zr 3 The Fe phase contains Cu, zr 2 The Cu phase contains Fe. It can be seen that trace Cu addition was specific to SThe distribution characteristics of PPs have a significant impact. Addition of trace Cu replaces part of Fe and enriches Zr (FeCr) 2 The phase further influences the precipitation kinetics, refines the size of the alloy, and further improves the corrosion resistance of the alloy. In addition, when the Cu content is too low (< 0.02 wt.%) coarse-sized o-Zr appears 3 A Fe phase; when the Cu content is too high (> 0.08 wt.%) a larger size of t-Zr is formed 2 A Cu phase. These second phases may reduce the corrosion resistance of the zirconium alloy, so the optimum content of Cu is between 0.02 and 0.08 wt.%.
Test example 4
FIG. 4 is a graph showing the corrosion weight gain of the zirconium alloys prepared in examples 1-2 and comparative examples 1-2 in a 300ppb dissolved oxygen high temperature environment (360 ℃ C., 20 MPa). As can be seen from fig. 4, the 0.07Cu zirconium alloy (example 2) has the best corrosion resistance, the initial corrosion rate is slower and the 240d weight gain is the lowest. The 0.05Cu zirconium alloy (example 1) has a corrosion law similar to that of 0.07Cu, but the weight gain rate is obviously higher than that of the 0.05Cu zirconium alloy in the corrosion turning (around 60-90 d), so that the 240d weight gain is higher. From the point of 240d weight gain, the corrosion resistance of the alloys according to examples 1-2 were better than commercial Zr-4 alloys. The 0.13Cu zirconium alloy (comparative example 1) has a faster initial corrosion rate, comparable to the reference Zr-4, M5 and ZIRLO alloys, and even though the corrosion turning time and post-turning corrosion rates are not much different from those of the 0.05Cu and 0.07Cu zirconium alloys, it results in a much higher final corrosion weight gain (240 d) than the zirconium alloys of examples 1 and 2, just inferior to the M5 alloy. The initial corrosion rate of the 0Cu zirconium alloy (comparative example 2) was the lowest, but the corrosion after turning increased faster, so the final weight gain was also higher, slightly lower than that of the 0.13Cu zirconium alloy. The alloys of comparative examples 1-2 were inferior in corrosion performance (240 d weight gain) to the Zr-4 alloy, with the 0.13Cu zirconium alloy being due to its higher initial corrosion rate and the 0Cu zirconium alloy being due to its higher post-inflection corrosion rate. In conclusion, the corrosion resistance of the 0.05Cu and 0.07Cu zirconium alloy in 300ppb dissolved oxygen water environment is superior to that of other comparative alloys and reference commercial alloys due to the lower initial corrosion rate and the corrosion rate after turning, so the alloy is suitable for oxygen-enriched high-temperature high-pressure water environment.
Test example 5
FIG. 5 is a graph showing the corrosion weight gain curves of the zirconium alloys prepared in examples 3-4 and comparative examples 3-4 in a 1000ppb dissolved oxygen high temperature environment (360 ℃ C., 20 MPa). As can be seen from FIG. 5, the 0.05Cu and 0.07Cu zirconium alloys (examples 3, 4) have similar corrosion laws, and although their initial corrosion rates are greatly improved relative to the 300ppb dissolved oxygen environment, comparable to that of the Zr-4 alloy, their post-inflection corrosion rates are slower, and the final (240 d) weight gain is only slightly higher than that of the Zr-4 alloy. Considering that the dissolved oxygen content of the practical small micro reactor working condition is difficult to reach 1000ppb and comprehensively considering the corrosion resistance of the environment of 300ppb, 0.05Cu and 0.07Cu zirconium alloy are considered as alternative alloys which are more suitable for the oxygen-enriched water environment of the small micro nuclear reactor. However, the 0.13Cu zirconium alloy (comparative example 3) has good corrosion performance due to its lower initial corrosion rate, and 240d has a lower corrosion weight gain than the Zr-4 alloy, so it is more suitable for high concentration dissolved oxygen water environments. The 0Cu zirconium alloy (comparative example 4) has high corrosion rate after turning due to the occurrence of two corrosion turns, and the final weight gain is equivalent to Nb-containing alloys such as M5, ZIRLO and the like, and is not suitable for oxygen-enriched water environment.
Test example 6
FIG. 6 is a graph showing the corrosion weight gain of the zirconium alloys prepared in examples 4 through 5 and comparative example 5 in a 1000ppb dissolved oxygen high temperature environment (360 ℃ C., 20 MPa). As can be seen from FIG. 6, the 0.07Cu zirconium alloy in the destressing state annealed at 470 ℃ has the best corrosion resistance, the initial corrosion rate is lower, and the weight gain of 240d is lower than that of the Zr-4 alloy. The initial corrosion rate of the 0.07Cu zirconium alloy in the partially recrystallized state annealed at 550 c is higher, so the final weight gain (240 d) is slightly higher than Zr-4. The partially recrystallized 0.07Cu alloy annealed at 510 ℃ has a corrosion weight gain law similar to that of the 550 ℃ annealed alloy, but has a corrosion weight gain higher than that of the 550 ℃ annealed alloy. Therefore, the 0.07Cu zirconium alloy subjected to stress relief annealing and partial recrystallization annealing has good corrosion resistance, and is suitable for oxygen-enriched water environment. It should be noted that in a high concentration of 1000ppb dissolved oxygen environment, a lower temperature stress relief anneal may be more beneficial to improve corrosion resistance than a partial recrystallization anneal, which may be related to higher matrix strength itself, SPPs distribution morphology, and grain size and orientation.
Test example 7
TABLE 2 Corrosion weight gain results (mg/dm) for zirconium alloys and reference commercial alloys in various oxygen-containing aqueous environments for various examples and comparative examples 2 )
| For 6 days | For 30 days | For 60 |
90 |
120 |
150 |
180 |
210 |
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 |
| Zr-4 300ppb | 11.87 | 19.52 | 34.33 | 39.32 | 40.68 | 40.87 | 43.29 | 45.27 | 58.59 |
| M5 300ppb | 12.18 | 20.54 | 33.74 | 44.39 | 54.37 | 62.97 | 71.82 | 81.77 | 83.45 |
| ZIRLO 300ppb | 12.76 | 22.49 | 33.24 | 41.63 | 45.47 | 48.50 | 54.58 | 62.58 | 68.87 |
| Zr-4 1000ppb | 11.44 | 21.24 | 34.64 | 33.93 | 36.32 | 40.18 | 42.95 | 48.58 | 59.67 |
| M5 1000ppb | 11.50 | 21.48 | 38.52 | 47.09 | 55.33 | 62.34 | 71.01 | 79.86 | 82.46 |
| ZIRLO 1000ppb | 12.08 | 28.84 | 46.87 | 51.00 | 56.15 | 62.28 | 68.13 | 76.40 | 82.62 |
As can be seen from Table 2, the zirconium alloy provided by the invention has excellent corrosion resistance in an oxygen-enriched water environment. This is consistent with the results of test examples 4 to 6 and FIGS. 4 to 6. Note that in table 2, the oxygen-containing water environments of examples 1 to 5 and comparative examples 1 to 5 are identical to table 1.
Test example 8
FIG. 7 is a Scanning Electron Microscope (SEM) cross-sectional fracture map (300 ppb) and a Transmission Electron Microscope (TEM) oxide film energy spectrum (EDS) chemical composition plane distribution map and a crystal structure analysis map (1000 ppb) of the 0.07Cu zirconium alloy prepared in examples 2, 4, 5 and comparative example 5 corroded 240d oxide film in high temperature and high pressure water. FIG. 7 (a)A cross-sectional macro-morphology graph of a 0.07Cu zirconium alloy oxide film, (a) 1 ) An enlarged view of the rectangular frame location area in (a) showing a transverse crack or oxide film "delamination"; (b) TEM morphology of 0.07Cu zirconium alloy, zr, O, fe, cr and EDS surface scan of Cu; (c) A cross-sectional TEM profile of the 0.07Cu zirconium alloy oxide film, showing an oxide film/matrix (O/M) interface and a transition layer (ZrO); (c) 1 )、(c 2 ) And (c) 3 ) For ZrO, zrO/alpha-Zr interface, zrO/ZrO in (c) 2 Is a high resolution TEM image of the interface of (a). As can be seen from fig. 7 (a), the oxide film of the zirconium alloy is completely bonded to the substrate, but the cross-sectional fracture has a certain "delamination" separated by transverse cracks, and this morphology confirms the presence of a large internal stress in the oxide film. Fig. 7 (b) shows the elemental distribution of the oxide film cross section, and the difference in component contrast between Zr and O distinguishes the oxide film and the α -Zr matrix, and the (O/M) interface between them fluctuates and fluctuates, and the oxide film above a portion of the peak is subject to tensile stress to generate transverse cracks. Both such interfacial fluctuations and transverse cracks are a result of the action of internal stress in the oxide film. The larger Fe and Cu element-enriched regions represent Zr respectively 3 Fe and Zr 2 The presence of Cu phases, both of which are only present in the α -Zr matrix region, suggests that both of these second phases may rapidly oxidize and dissolve to the surrounding oxide film during the etching process. The fine Fe and Cr element rich regions represent Zr (FeCr) 2 The phases are dispersed in the alpha-Zr matrix and the oxide film area, and the second phase delays oxidation in the corrosion process and stays in the oxide film with ceramic property in the form of 'particles' with metal property, so that the internal stress of the oxide film is relieved. It should be noted that in the oxide film region far from the O/M interface, the enriched region of Cr is significantly more than Fe, which indicates that when Zr (FeCr) 2 During final oxidation of the phase, fe element diffuses to the surrounding matrix, realizing Fe 2+ Or Fe (Fe) 3+ Ion pair ZrO 2 The "doping" of the crystal suppresses the generation of oxygen vacancies, thereby reducing the diffusion of aggressive media. In addition, these fine and dispersed Zr (FeCr) 2 During oxidation of the phase, the local volume expansion of the oxide film introduces less stress variation, so that the stress is more uniformly distributed to reduce the generation of micro-voids or microcracks, therebyThe compactness and the protection of the oxide film are improved. FIG. 7 (c) shows h-ZrO of hexagonal structure existing near the O/M interface as a transition layer between the α -Zr matrix and the oxide film, which not only relieves the phase transition stress during oxidation but also reduces the rate of corrosion reaction. When a certain amount of dissolved oxygen exists in the corrosion medium or the alpha-Zr matrix contains trace O elements, the morphological characteristics of the interface transition layer can be influenced, and the corrosion reaction process is further influenced. In summary, the microstructure characteristics, such as finely dispersed Zr (FeCr), are determined by the chemical composition and the preparation process 2 The phase, relatively larger grain size or smaller orientation difference and the O/M interface transition layer can relieve the internal stress of the oxide film, improve the capability of the oxide film for preventing the diffusion of the aggressive medium, and further improve the corrosion resistance.
The foregoing is merely a preferred embodiment of the present invention and it should be noted that modifications and adaptations to those skilled in the art may be made without departing from the principles of the present invention, which are intended to be comprehended within the scope of the present invention.
Claims (10)
1. A zirconium alloy for a nuclear reactor, comprising, in weight percent: 0.40 to 0.65 percent of Sn, 0.12 to 0.25 percent of Nb, 0.35 to 0.50 percent of Fe, 0.15 to 0.20 percent of Cr, 0 to 0.13 percent of Cu, 0.08 to 0.16 percent of O and the balance of Zr.
2. The zirconium alloy for nuclear reactor of claim 1, wherein the zirconium alloy comprises an a-Zr matrix and second phase precipitate particles; the alpha-Zr matrix is of a close-packed hexagonal structure; the second phase precipitation particles include Zr (FeCr) 2 Phase, o-Zr 3 Fe phase and t-Zr 2 A Cu phase.
3. The zirconium alloy for nuclear reactor of claim 2, wherein the α -Zr matrix has an average grain size between 1.02±1.12 to 7.57±3.47 μm, is a destressing or partially recrystallized microstructure, and has a recrystallization fraction between 13 and 78%; the second phase precipitated particles are spherical or spheroidal and are dispersed or strip-shaped and distributed in the crystal and grain boundary of the alpha-Zr matrix, and the average size is between 25 and 420 nm.
4. A method for preparing the zirconium alloy for nuclear reactor according to any one of claims 1 to 3, comprising the steps of:
smelting a metal raw material to obtain a zirconium alloy cast ingot; the components of the zirconium alloy cast ingot are consistent with the chemical components of the zirconium alloy in any one of claims 1-3;
performing hot press molding or forging molding on the zirconium alloy cast ingot to obtain a blank;
and sequentially carrying out homogenization treatment, hot rolling, cold rolling and annealing treatment on the blank to obtain the zirconium alloy for the nuclear reactor.
5. The method according to claim 4, wherein the hot press molding temperature is 660 to 720 ℃; the forging forming temperature is 920-1120 ℃.
6. The method according to claim 4, wherein the homogenization treatment is performed at a temperature of 1000 to 1070 ℃ for a time of 0.5 to 4 hours, and then quenched to room temperature.
7. The method according to claim 4, wherein the hot rolling is carried out at a temperature of 620 to 750 ℃.
8. The method according to claim 4, wherein the cold rolling is multi-pass rolling, the reduction of each pass is 18-45%, the intermediate annealing temperature between passes is 480-600 ℃, and the heat preservation time is 1-6 h.
9. The method according to claim 4, wherein the annealing treatment is performed at a temperature of 450 to 600 ℃ for 3 to 7 hours; the atmosphere of the annealing treatment is vacuum.
10. Use of the zirconium alloy for nuclear reactor according to any one of claims 1 to 3 or the zirconium alloy for nuclear reactor according to any one of claims 4 to 9 in small miniature water-cooled nuclear reactor.
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