CN114703397B - Zirconium-based alloy with corrosion resistance and creep resistance for nuclear reactor fuel cladding and method for preparing zirconium-based alloy pipe by using zirconium-based alloy - Google Patents
Zirconium-based alloy with corrosion resistance and creep resistance for nuclear reactor fuel cladding and method for preparing zirconium-based alloy pipe by using zirconium-based alloy Download PDFInfo
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- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
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- C22C16/00—Alloys based on zirconium
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- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
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- C22C1/00—Making non-ferrous alloys
- C22C1/02—Making non-ferrous alloys by melting
- C22C1/03—Making non-ferrous alloys by melting using master alloys
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- C—CHEMISTRY; METALLURGY
- 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
- C22F1/002—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working by rapid cooling or quenching; cooling agents used therefor
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- C—CHEMISTRY; METALLURGY
- 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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- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
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Abstract
The application discloses a zirconium-based alloy with corrosion resistance and creep resistance for nuclear reactor fuel cladding and a method for preparing a zirconium-based alloy pipe by using the zirconium-based alloy, wherein the zirconium-based alloy comprises the following components in percentage by mass: nb 0.2-2.0%, V0.005-0.11%, fe 0.014-0.5%, O0.08-0.16%, S0.0015-0.03%, C not more than 70ppm, si not more than 70ppm, ti not more than 40ppm, N not more than 35ppm, and the mass ratio of V to Fe in the zirconium-based alloy is 1: (2-3), the balance being Zr and unavoidable impurities. The pipe prepared from the zirconium-based alloy has excellent corrosion resistance and creep property through practical verification.
Description
Technical Field
The application belongs to the technical field of zirconium-based alloy materials of structural materials for nuclear reactors, and particularly relates to a zirconium-based alloy for nuclear reactor fuel cladding with corrosion resistance and creep property and a method for preparing a zirconium-based alloy pipe by using the zirconium-based alloy.
Background
The nuclear energy is taken as an important component of a modern energy system, the energy development is brought into important content of ecological civilization construction in a report of 'ninety-nine-big', the long-level conference of the clean energy department of the ninth world in 2018 definitely brings the nuclear energy into the clean energy, the high-quality development of the nuclear energy in China is in strategic opportunity, and the bottleneck of the development of the nuclear energy-the autonomous research of zirconium materials for nuclear power is also urgent.
Zirconium and its alloys are widely used as structural components and fuel cladding in nuclear power reactors, mainly because zirconium has a low thermal neutron absorption cross section, strong corrosion resistance in high temperature water and high mechanical strength. In a nuclear reactor, improving the burnup of nuclear fuel is an effective way to reduce nuclear power costs, while for pressurized water reactors, the main limiting factors for further improving burnup are water side corrosion, hydrogen absorption and excellent mechanical properties of fuel cladding zirconium alloys. Thus, this places higher demands on the corrosion resistance of the zirconium alloy.
Research has shown that zirconium alloys for use in the nuclear industry now and in the future will be predominantly tertiary and minor amounts of secondary modified zirconium alloys and zirconium alloy demand is expected to exceed 3000 tons per year. The M5 zirconium alloy is widely used in the third generation nuclear power unit due to its excellent processability and nuclear performance. Therefore, research on the autonomous zirconium alloy suitable for industrialization is urgent, and the application aims to find out elements suitable for industrial production without reducing the excellent performance of Zr-Nb series alloys by researching the influence of different elements on the performance of the zirconium-niobium alloy, thereby further developing the autonomous zirconium alloy. With reference to the development trend of the world zirconium alloy, the principle of multi-element small-amount zirconium alloying is inherited, and the adjustment of different proportions of alloy elements or the addition of other kinds of alloy elements on the basis of the existing alloy components is the correct direction for developing novel zirconium alloy.
Based on the logic, the zirconium-based alloy (patent number is CN 105018795B) for nuclear reactor fuel cladding with excellent corrosion resistance, which is developed before the research team, well solves the corrosion problem of the zirconium-based alloy in-reactor reaction, but has poor creep property. When the zirconium alloy is used as a cladding material of a fuel assembly and is in long-term service under a high-temperature high-pressure irradiation environment, the critical creep rate of the material is exceeded, the material can creep, cracking can be caused near an oxide/metal interface, the cracking can be inwards extended to reach the metal surface, and the material is cracked, so that irradiation accidents are caused. Therefore, the zirconium-based alloy cannot fully meet the requirement of high burnup of the next generation pressurized water reactor, and the research team aims to develop a zirconium-based alloy which can simultaneously have excellent corrosion resistance and excellent creep property through further research on the zirconium-based alloy.
Disclosure of Invention
The application aims to overcome the defects in the prior art and provide a zirconium-based alloy for nuclear reactor fuel cladding with corrosion resistance and creep property and a method for preparing a zirconium-based alloy pipe by immersing the alloy in deionized water, wherein the corrosion weight gain of the zirconium-based alloy is not more than 70mg & dm after 200 days of corrosion under the conditions of the temperature of 360 ℃ and the pressure of 18.6MPa -2 The method comprises the steps of carrying out a first treatment on the surface of the The alloy is put into deionized water vapor atmosphere, the temperature is 500 ℃, the corrosion weight gain is not more than 310mg dm after corrosion for 500 hours under the pressure of 10.3MPa -2 The method comprises the steps of carrying out a first treatment on the surface of the The zirconium-based alloy has excellent creep property, so that the zirconium-based alloy has excellent mechanical property and processing property, the surface is respectively 100MPa, 117MPa, 137MPa and 157MPa at 400 ℃, the uniaxial tensile creep test time is longer than 200h, the pressure is respectively 117MPa, 137MPa and 157MPa at 350 ℃, the uniaxial tensile creep test time is longer than 200h, the pressure is respectively 117MPa, 137MPa and 157MPa at 300 ℃, and the uniaxial tensile creep test time is longer than 200h.
The application aims at solving the problems by the following technical scheme:
a zirconium-based alloy for nuclear reactor fuel cladding having both corrosion and creep properties, the zirconium-based alloy comprising the following components in mass percent: nb 0.2-2.0%, V0.005-0.11%, fe 0.014-0.5%, O0.08-0.16%, S0.0015-0.03%, C not more than 70ppm, si not more than 70ppm, ti not more than 40ppm, N not more than 35ppm, and the mass ratio of V to Fe in the zirconium-based alloy is 1: (2-3), the balance being Zr and unavoidable impurities;
corrosion resistance of the zirconium based alloyThe etching performance is shown as follows: immersing the zirconium-based alloy into deionized water, and corroding for 200 days under the conditions of the temperature of 360 ℃ and the pressure of 18.6MPa, wherein the corrosion weight gain of the zirconium-based alloy is less than or equal to 70mg & dm -2 The method comprises the steps of carrying out a first treatment on the surface of the Placing the zirconium-based alloy in deionized water steam atmosphere, and corroding for 500 hours at the temperature of 500 ℃ and the pressure of 10.3MPa, wherein the corrosion weight gain of the zirconium-based alloy is less than or equal to 310mg dm -2 The method comprises the steps of carrying out a first treatment on the surface of the The creep properties of the zirconium based alloy are shown as follows: the pressures are respectively 100MPa, 117MPa, 137MPa and 157MPa at 400 ℃, the uniaxial tensile creep test time is longer than 200h, the pressures are respectively 117MPa, 137MPa and 157MPa at 350 ℃, the uniaxial tensile creep test time is longer than 200h, and the pressures are respectively 117MPa, 137MPa and 157MPa at 300 ℃, and the uniaxial tensile creep test time is longer than 200h.
Further, the raw materials of the zirconium-based alloy comprise nuclear sponge zirconium, ferrovanadium zirconium, niobium zirconium alloy and ferrous sulfide or elemental sulfur.
Further, the zirconium-based alloy consists of the following components in percentage by mass: nb 0.2%, V0.05%, fe 0.5%, O0.16%, S0.002%, C70 ppm or less, si 70ppm or less, ti 40ppm or less, N35 ppm or less, and Zr and unavoidable impurities in balance. Further, the zirconium-based alloy consists of the following components in percentage by mass: nb is 0.9%, V is 0.11%, fe is 0.2%, O is 0.09%, S is 0.003%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
Further, the zirconium-based alloy consists of the following components in percentage by mass: nb 2.0%, V0.07%, fe 0.15%, O0.14%, S0.0015%, C70 ppm or less, si 70ppm or less, ti 40ppm or less, N35 ppm or less, and Zr and unavoidable impurities as the rest. Further, the zirconium-based alloy consists of the following components in percentage by mass: nb is 0.9%, V is 0.01%, fe is 0.03%, O is 0.16%, S is 0.0035%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
Further, the zirconium-based alloy consists of the following components in percentage by mass: nb is 0.8%, V is 0.04%, fe is 0.12%, O is 0.08%, S is 0.03%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities. Further, the zirconium-based alloy consists of the following components in percentage by mass: nb is 1.1%, V is 0.005%, fe is 0.014%, O is 0.09%, S is 0.007%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
The preparation method of the zirconium-based alloy pipe selects the zirconium-based alloy as a base for preparation, and specifically comprises the following steps:
selecting raw materials according to the components of the zirconium-based alloy, and smelting the raw materials into a zirconium-iron-vanadium intermediate alloy and a zirconium-niobium intermediate alloy;
preparing an alloy bag from zirconium-iron-vanadium alloy and ferrous sulfide or elemental sulfur by using a zirconium foil, uniformly placing the alloy bag in the middle of zirconium sponge, pressing the alloy bag into an electrode, and carrying out conventional smelting for three times in a vacuum consumable arc furnace to obtain an alloy cast ingot;
step three, cogging and forging the alloy ingot obtained in the step two at the forging temperature of 960-1070 ℃ to obtain a bar blank;
and fourthly, carrying out solution treatment on the rod blank obtained in the third step at 960-1070 ℃, quenching, extruding the rod blank into a tube blank at 600-700 ℃, and carrying out multi-pass cold rolling, intermediate annealing and finished product annealing to finally obtain the zirconium-based alloy tube.
Further, the pass deformation of the cold rolling in the fourth step is 45% -85%, the intermediate annealing temperature is 560-650 ℃, the finished product annealing temperature is 540-600 ℃, the time is 2-5 h, and the vacuum degree is less than or equal to 8.0X10 -2 Under Pa.
Compared with the prior art, the application has the following beneficial effects:
1. the solubility of sulfur in alpha-Zr in the zirconium based alloy of the present application is relatively low and when S is added in excess of 100 mug/g in pure Zr, a second phase containing S may occur. The intermetallic compound formed by adding sulfur into the zirconium alloy contains Zr 9 S 2 Zr generated in the niobium-containing zirconium-based alloy with tetragonal structure 9 S 2 The intermediate alloy and S dissolved in zirconium improve creep property of the alloy, but the excessively high sulfur content can increase the thermal neutron absorption section of the zirconium alloy, and influence the service efficiency of the zirconium alloy under irradiation conditions. Therefore, the application adds a certain amount of S in the Zr-Nb-Fe-V alloy, which not only can improve the corrosion performance of the alloy, but also can improve the creep property of the alloy and improve the mechanical property of the zirconium-niobium alloy. And through practical verification, the corrosion resistance of the zirconium-based alloy of the application is shown as follows: immersing a zirconium-based alloy in deionized water, and corroding for 200 days at the temperature of 360 ℃ and the pressure of 18.6MPa, wherein the corrosion weight gain of the zirconium-based alloy is less than or equal to 70mg & dm -2 The method comprises the steps of carrying out a first treatment on the surface of the Placing the zirconium-based alloy in deionized water steam atmosphere, and corroding for 500 hours at the temperature of 500 ℃ and the pressure of 10.3MPa, wherein the corrosion weight gain of the zirconium-based alloy is less than or equal to 310mg & dm -2 The method comprises the steps of carrying out a first treatment on the surface of the The creep properties of the zirconium based alloy are also shown as: the pressures are respectively 100MPa, 117MPa, 137MPa and 157MPa at 400 ℃, the uniaxial tensile creep test time is longer than 200h, the pressures are respectively 117MPa, 137MPa and 157MPa at 350 ℃, the uniaxial tensile creep test time is longer than 200h, and the pressures are respectively 117MPa, 137MPa and 157MPa at 300 ℃, and the uniaxial tensile creep test time is longer than 200h.
2. The application limits the content of titanium (Ti) in the zirconium-based alloy, ti is an infinite miscible solid solution in Zr, ti can reduce the alpha-beta transition temperature of the alloy, not only affects the corrosion resistance of the zirconium alloy, but also reduces the processing window of the zirconium alloy. Therefore, the application particularly limits the content of Ti element in the Zr-Nb-Fe-V alloy, and can improve the processing window under the condition of ensuring the corrosion performance of the alloy.
3. The application limits the content of silicon (Si) in the zirconium-based alloy, and the solubility of Si in Zr is very low, thus forming Zr 4 Si、Zr 2 Si、Zr 3 Si 2 、Zr 4 Si 3 、Zr 6 Si 5 、ZrSi、ZrSi 2 The second phase is not advantageous for corrosion, in order to avoid the formation of the above mesophases. Therefore, in order to ensure the high corrosion resistance of the alloy of the application, the applicationThe application particularly limits the content of Si element in Zr-Nb-Fe-V alloy.
4. The content of carbon (C) and nitrogen (N) in the zirconium-based alloy is limited, and for Zr alloys, C and N are both harmful elements. Therefore, in order to ensure the high corrosion resistance of the alloy of the application, the application particularly limits the content of C, N element in the Zr-Nb-Fe-V alloy.
5. The zirconium-based alloy is a zirconium-niobium alloy containing iron, vanadium and sulfur, wherein niobium is a beta-phase stabilizing element, niobium has higher strengthening effect on zirconium, and researches show that when a small amount of niobium is added, the corrosion resistance of a zirconium alloy material can be improved, but the niobium content is too high and is sensitive to heat treatment, so that the adding amount of niobium in the zirconium-based alloy is not more than 2wt%, and the zirconium-based alloy can be ensured to have excellent corrosion resistance and good mechanical property; in addition, the oxygen element can form an interstitial solid solution in the zirconium-based alloy, has larger solubility and has obvious strengthening effect. The interstitial solid solution can improve the mechanical strength of the alloy, but the too low oxygen content has unobvious improving effect, can not meet the required performance requirement, and the too high oxygen content can reduce the workability of the alloy, and the zirconium-niobium alloy has the oxygen content of 0.06-0.16 wt percent, so that the mechanical strength of the alloy can be improved and the good workability of the alloy can be maintained.
6. The iron in the zirconium-based alloy of the application can reduce the alpha-beta transformation temperature of the alloy, the solubility of the iron in alpha-Zr is about 0.02 percent, the maximum solubility in beta-Zr is 5.5 percent, the magnetic transformation temperature is 769 ℃ after pure iron is added into the zirconium alloy, and the formed intermetallic compound has Zr 2 Fe and ZrFe 2 Wherein ZrFe 2 With C15 (MgCu) 2 ) Type structure of (Zr, nb) Fe formed in niobium-containing zirconium-based alloy 2 The intermediate alloy improves the corrosion performance of the alloy, but the too high content of iron can influence the processing performance of the alloy and the yield of products, so that the application adds a small amount of V in the Zr-Nb-Fe alloy and can not influence the processing performance of the alloy under the condition of improving the corrosion performance of the alloy; vanadium is usually an impurity in zirconium alloys, the content of which is controlled to be less than 0.005wt%, but vanadium can reduce the compositionThe alpha-beta transformation temperature of gold, in the application, vanadium is added into the zirconium alloy as an alloy element, so that the processing performance of the alloy can be optimized and the mechanical property of the zirconium-niobium alloy can be improved under the condition of improving the excellent corrosion resistance of the alloy.
Detailed Description
The present application will be described in further detail with reference to examples for better understanding of the technical aspects of the present application by those skilled in the art.
The application provides a zirconium-based alloy with corrosion resistance and creep resistance for nuclear reactor fuel cladding, which comprises the following components in percentage by mass: nb is 0.2-2.0%, V is 0.005-0.5%, fe is 0.005-0.5%, O is 0.06-0.16%, S is 0.001-0.03%, C is less than or equal to 70ppm, si is less than or equal to 70ppm, ti is less than or equal to 40ppm, N is less than or equal to 35ppm, and the mass ratio of V to Fe in the zirconium-based alloy is 1: (2-3), the balance being Zr and unavoidable impurities; wherein the content of harmful elements C and N and the content of Si are defined in the zirconium-based alloy, so that Zr is prevented from being formed in the alloy 4 Si、Zr 2 The second phase such as Si affects the corrosion resistance of the alloy, and the content of Ti is limited, because Ti can reduce the alpha-beta transformation temperature of the alloy, the corrosion resistance of the zirconium alloy is affected, and meanwhile, the processing window of the zirconium alloy is reduced; meanwhile, a small amount of S is added, so that the corrosion performance of the alloy can be improved, the creep property of the alloy can be improved, and the mechanical property of the zirconium-niobium alloy can be improved.
Specifically, the raw materials for preparing the zirconium-based alloy comprise nuclear sponge zirconium, ferrovanadium alloy, zirconium niobium alloy and ferrous sulfide or elemental sulfur.
In addition, the application provides a preparation method of the zirconium-based alloy pipe according to the zirconium-based alloy, which specifically comprises the following steps:
selecting raw materials according to the components of the zirconium-based alloy, and smelting the raw materials into a zirconium-iron-vanadium intermediate alloy and a zirconium-niobium intermediate alloy;
preparing an alloy bag from zirconium-iron-vanadium alloy and ferrous sulfide or elemental sulfur by using a zirconium foil, uniformly placing the alloy bag in the middle of zirconium sponge, pressing the alloy bag into an electrode, and carrying out conventional smelting for three times in a vacuum consumable arc furnace to obtain an alloy cast ingot;
step three, cogging and forging the alloy ingot obtained in the step two at the forging temperature of 960-1070 ℃ to obtain a bar blank;
and fourthly, carrying out solution treatment on the rod blank obtained in the third step at the temperature of 1000-1070 ℃, extruding the rod blank into a tube blank at the temperature of 600-700 ℃ after quenching, and finally obtaining the zirconium-based alloy tube after multi-pass cold rolling, intermediate annealing and finished product annealing.
Wherein the pass deformation of the cold rolling in the fourth step is 45% -85%, the intermediate annealing temperature is 560-650 ℃, the finished product annealing temperature is 540-600 ℃, the time is 2-5 h, and the vacuum degree is less than or equal to 8.0X10 -2 Under Pa.
In order to verify the efficacy of the application in preparing zirconium based alloys, the application was tested in the following manner:
example 1
The zirconium-based alloy for the nuclear reactor fuel cladding with corrosion resistance and creep property comprises the following components in percentage by mass: nb 0.2%, V0.05%, fe 0.5%, O0.16%, S0.002%, C70 ppm or less, si 70ppm or less, ti 40ppm or less, N35 ppm or less, and Zr and unavoidable impurities in balance.
The preparation method of the zirconium based alloy pipe comprises the following steps: firstly, selecting raw materials according to the design components of a zirconium-based alloy, and smelting the raw materials into a zirconium-iron-vanadium intermediate alloy and a zirconium-niobium intermediate alloy; then using zirconium foil to prepare alloy bags from ferrovanadium zirconium and ferrous sulfide or elemental sulfur, uniformly placing the alloy bags in the middle of sponge zirconium, pressing the sponge zirconium into electrodes, and carrying out three times of conventional smelting in a vacuum consumable arc furnace to obtain alloy ingots; then, cogging and forging the alloy ingot at 980 ℃ to obtain a bar blank, carrying out solid solution treatment at 1070 ℃ to the bar blank, quenching, extruding the bar blank into a tube blank at 650 ℃, and carrying out multi-pass cold rolling, intermediate annealing and finished product annealing to finally obtain the zirconium-based alloy tubeThe method comprises the steps of carrying out a first treatment on the surface of the The pass deformation of the cold rolling is 45-85%, the intermediate temperature is 600 ℃, the finished product annealing temperature is 560 ℃, the time is 2-5 h, and the vacuum degree is not more than 8.0X10 -2 And (3) under the Pa condition, finally obtaining the zirconium-based alloy pipe with the diameter of 9.5mm multiplied by 0.57mm multiplied by Lmm.
Example 2
The zirconium-based alloy for the nuclear reactor fuel cladding with corrosion resistance and creep property comprises the following components in percentage by mass: nb is 2.0%, V is 0.11%, fe is 0.2%, O is 0.09%, S is 0.003%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
The preparation method of the zirconium based alloy pipe comprises the following steps: firstly, selecting raw materials according to the design components of a zirconium-based alloy, and smelting the raw materials into a zirconium-iron-vanadium intermediate alloy and a zirconium-niobium intermediate alloy; then using zirconium foil to prepare alloy bags from ferrovanadium zirconium and ferrous sulfide or elemental sulfur, uniformly placing the alloy bags in the middle of sponge zirconium, pressing the sponge zirconium into electrodes, and carrying out three times of conventional smelting in a vacuum consumable arc furnace to obtain alloy ingots; then, cogging and forging the alloy ingot at 960 ℃ to obtain a bar blank, carrying out solid solution treatment on the bar blank at 1070 ℃, extruding the bar blank into a tube blank at 620 ℃ after quenching, and carrying out multi-pass cold rolling, intermediate annealing and finished product annealing to finally obtain the zirconium-based alloy tube; the pass deformation of the cold rolling is 45-85%, the intermediate temperature is 560 ℃, the annealing temperature of the finished product is 540 ℃, the time is 2-5 h, and the vacuum degree is not more than 8.0X10 -2 And (3) under the Pa condition, finally obtaining the zirconium-based alloy pipe with the diameter of 9.5mm multiplied by 0.57mm multiplied by Lmm.
Example 3
The zirconium-based alloy for the nuclear reactor fuel cladding with corrosion resistance and creep property comprises the following components in percentage by mass: nb is 1.2%, V is 0.07%, fe is 0.15%, O is 0.14%, S is 0.0015%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
Zirconium based alloy of the present exampleThe preparation method of the pipe comprises the following steps: firstly, selecting raw materials according to the design components of a zirconium-based alloy, and smelting the raw materials into a zirconium-iron-vanadium intermediate alloy and a zirconium-niobium intermediate alloy; then using zirconium foil to prepare alloy bags from ferrovanadium zirconium and ferrous sulfide or elemental sulfur, uniformly placing the alloy bags in the middle of sponge zirconium, pressing the sponge zirconium into electrodes, and carrying out three times of conventional smelting in a vacuum consumable arc furnace to obtain alloy ingots; then, cogging and forging the alloy ingot at the forging temperature of 1050 ℃ to obtain a bar blank, carrying out solid solution treatment on the bar blank at the forging temperature of 1050 ℃, extruding the bar blank into a tube blank at the temperature of 620 ℃ after quenching, and carrying out multi-pass cold rolling, intermediate annealing and finished product annealing procedures to finally obtain the zirconium-based alloy tube; the pass deformation of the cold rolling is 45-85%, the intermediate temperature is 650 ℃, the annealing temperature of the finished product is 600 ℃, the time is 2-5 h, and the vacuum degree is not more than 8.0X10 -2 And (3) under the Pa condition, finally obtaining the zirconium-based alloy pipe with the diameter of 9.5mm multiplied by 0.57mm multiplied by Lmm. .
Example 4
The zirconium-based alloy for the nuclear reactor fuel cladding with corrosion resistance and creep property comprises the following components in percentage by mass: nb is 0.9%, V is 0.01%, fe is 0.03%, O is 0.16%, S is 0.0035%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
The preparation method of the zirconium based alloy pipe comprises the following steps: firstly, selecting raw materials according to the design components of a zirconium-based alloy, and smelting the raw materials into a zirconium-iron-vanadium intermediate alloy and a zirconium-niobium intermediate alloy; then using zirconium foil to prepare alloy bags from ferrovanadium zirconium and ferrous sulfide or elemental sulfur, uniformly placing the alloy bags in the middle of sponge zirconium, pressing the sponge zirconium into electrodes, and carrying out three times of conventional smelting in a vacuum consumable arc furnace to obtain alloy ingots; then, cogging and forging the alloy ingot at 980 ℃ to obtain a bar blank, carrying out solid solution treatment at 1070 ℃ to the bar blank, quenching, extruding the bar blank into a tube blank at 650 ℃, and carrying out multi-pass cold rolling, intermediate annealing and finished product annealing to finally obtain the zirconium-based alloy tubeThe method comprises the steps of carrying out a first treatment on the surface of the The pass deformation of the cold rolling is 45-85%, the intermediate temperature is 580 ℃, the annealing temperature of the finished product is 540 ℃, the time is 2-5 h, and the vacuum degree is not more than 8.0X10 -2 And (3) under the Pa condition, finally obtaining the zirconium-based alloy pipe with the diameter of 9.5mm multiplied by 0.57mm multiplied by Lmm.
Example 5
The zirconium-based alloy for the nuclear reactor fuel cladding with corrosion resistance and creep property comprises the following components in percentage by mass: nb is 0.8%, V is 0.04%, fe is 0.12%, O is 0.08%, S is 0.03%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
The preparation method of the zirconium based alloy pipe comprises the following steps: firstly, selecting raw materials according to the design components of a zirconium-based alloy, and smelting the raw materials into a zirconium-iron-vanadium intermediate alloy and a zirconium-niobium intermediate alloy; then using zirconium foil to prepare alloy bags from ferrovanadium zirconium and ferrous sulfide or elemental sulfur, uniformly placing the alloy bags in the middle of sponge zirconium, pressing the sponge zirconium into electrodes, and carrying out three times of conventional smelting in a vacuum consumable arc furnace to obtain alloy ingots; then, cogging and forging the alloy ingot at 980 ℃ to obtain a bar blank, carrying out solid solution treatment on the bar blank at 1070 ℃, extruding the bar blank into a tube blank at 700 ℃ after quenching, and carrying out multiple cold rolling, intermediate annealing and finished product annealing procedures to finally obtain the zirconium-based alloy tube; the pass deformation of the cold rolling is 45-85%, the intermediate temperature is 600 ℃, the finished product annealing temperature is 560 ℃, the time is 2-5 h, and the vacuum degree is not more than 8.0X10 -2 And (3) under the Pa condition, finally obtaining the zirconium-based alloy pipe with the diameter of 9.5mm multiplied by 0.57mm multiplied by Lmm.
Example 6
The zirconium-based alloy for the nuclear reactor fuel cladding with corrosion resistance and creep property comprises the following components in percentage by mass: nb is 1.1%, V is 0.005%, fe is 0.014%, O is 0.09%, S is 0.007%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
Zirconium based alloy of the present exampleThe preparation method of the pipe comprises the following steps: firstly, selecting raw materials according to the design components of a zirconium-based alloy, and smelting the raw materials into a zirconium-iron-vanadium intermediate alloy and a zirconium-niobium intermediate alloy; then using zirconium foil to prepare alloy bags from ferrovanadium zirconium and ferrous sulfide or elemental sulfur, uniformly placing the alloy bags in the middle of sponge zirconium, pressing the sponge zirconium into electrodes, and carrying out three times of conventional smelting in a vacuum consumable arc furnace to obtain alloy ingots; then, cogging and forging the alloy ingot at 1070 ℃ to obtain a bar blank, carrying out solid solution treatment on the bar blank at 1000 ℃, extruding the bar blank into a tube blank at 620 ℃ after quenching, and carrying out multi-pass cold rolling, intermediate annealing and finished product annealing to finally obtain the zirconium-based alloy tube; the pass deformation of the cold rolling is 45-85%, the intermediate temperature is 580 ℃, the annealing temperature of the finished product is 580 ℃, the time is 2-5 h, and the vacuum degree is not more than 8.0X10 -2 And (3) under the Pa condition, finally obtaining the zirconium-based alloy pipe with the diameter of 9.5mm multiplied by 0.57mm multiplied by Lmm.
Finally, the corrosion performance of Zr-4 alloy (prepared by northwest nonferrous metal research institute) and zirconium-based alloy pipe for nuclear reactor fuel cladding prepared in examples 1 to 6 of the application were tested according to the national standard ASTM G2/G2M-2006, "test method for corrosiveness test of products of zirconium, hafnium and alloys thereof in 680 DEG F [360℃ ] water or 750 DEG F [400℃ ] steam", respectively, the test method being: the Zr-4 alloy and the zirconium-based alloy pipe for the nuclear reactor fuel cladding prepared in the examples 1 to 6 of the application are respectively placed in an autoclave, and after corrosion treatment, the corrosion weight gain is weighed; the conditions of the corrosion treatment are as follows: immersing the steel plate into deionized water at 360 ℃ and 18.6MPa for corrosion for 200 days; and placing the steel plate in a deionized water steam atmosphere to corrode for 500 hours at a temperature of 500 ℃ and a pressure of 10.3 MPa. The results of corrosion performance tests for Zr-4 alloys and zirconium-based alloys for nuclear reactor fuel cladding prepared in examples 1 to 6 of the present application are shown in Table 1.
TABLE 1 results of zirconium based alloy corrosion Performance test
As can be seen from Table 1, the zirconium based alloys for nuclear reactor fuel cladding prepared in examples 1 to 6 of the present application have excellent corrosion resistance at 360 ℃/18.6 MPa/deionized water/200 days and 500 ℃/10.3 MPa/deionized water vapor/500 hours as compared with the Zr-4 alloy, wherein the corrosion weight gain of the zirconium based alloys for nuclear reactor fuel cladding of examples 1 to 6 is not more than 65 mg. Dm under 360 ℃/18.6 MPa/deionized water/200 days -2 The method comprises the steps of carrying out a first treatment on the surface of the The zirconium based alloys for nuclear reactor fuel cladding of examples 1-6 had a corrosion weight gain of no more than 300 mg/dm at 500 ℃/10.3 MPa/deionized water vapor/500 hours -2 Far less than the corrosion weight gain of Zr-4 alloy under the same condition.
In addition, mechanical properties of the M5 alloy and zirconium-based alloys for nuclear reactor fuel cladding prepared in examples 1 to 6 of the present application were respectively tested, and the test results are shown in Table 2.
TABLE 2 results of zirconium based alloy mechanical Properties test
As can be seen from Table 2, the tensile strength and yield strength of the zirconium-based alloy for nuclear reactor fuel cladding of examples 1-6 of the present application are slightly better than the mechanical properties of Zr-4 alloy, which indicates that the addition of vanadium as an alloying element to the zirconium alloy of the present application not only improves the corrosion resistance of the zirconium alloy, but also optimizes the processability thereof and improves the mechanical properties of the zirconium-niobium alloy, and the zirconium-based alloy of the present application can be used as a nuclear reactor fuel cladding material or a structural material of a nuclear reactor.
Table 3 creep test data of 400 ℃/157Mpa
As can be seen from Table 3, the creep properties of the zirconium based alloys for nuclear reactor fuel cladding of examples 1 to 6 of the present application are all superior to those of the M5 alloy, showing that the addition of sulfur as an alloying element to the zirconium alloy of the present application not only improves the corrosion resistance of the zirconium alloy, but also optimizes the processability thereof, improves the creep properties of the zirconium-niobium alloy, and the zirconium based alloy of the present application can be used as a nuclear reactor fuel cladding material or a structural material of a nuclear reactor
The foregoing is only a specific embodiment of the application to enable those skilled in the art to understand or practice the application. Various modifications to these embodiments will be readily apparent to those skilled in the art, and the generic principles defined herein may be applied to other embodiments without departing from the spirit or scope of the application.
It will be understood that the application is not limited to what has been described above and that various modifications and changes may be made without departing from the scope thereof. The scope of the application is limited only by the appended claims.
Claims (10)
1. The zirconium-based alloy for the nuclear reactor fuel cladding with corrosion resistance and creep property is characterized by comprising the following components in percentage by mass: nb 0.2-2.0%, V0.005-0.11%, fe 0.014-0.5%, O0.08-0.16%, S0.0015-0.03%, C not more than 70ppm, si not more than 70ppm, ti not more than 40ppm, N not more than 35ppm, and the mass ratio of V to Fe in the zirconium-based alloy is 1: (2-3), the balance being Zr and unavoidable impurities;
the corrosion resistance of the zirconium based alloy is shown as follows: immersing the zirconium-based alloy into deionized water, and corroding for 200 days under the conditions of the temperature of 360 ℃ and the pressure of 18.6MPa, wherein the corrosion weight gain of the zirconium-based alloy is less than or equal to 70mg & dm -2 The method comprises the steps of carrying out a first treatment on the surface of the Placing the zirconium-based alloy in deionized water steam atmosphere, and corroding for 500 hours at the temperature of 500 ℃ and the pressure of 10.3MPa, wherein the corrosion weight gain of the zirconium-based alloy is less than or equal toAt 310 mg/dm -2 The method comprises the steps of carrying out a first treatment on the surface of the The creep properties of the zirconium based alloy are shown as follows: the pressures are respectively 100MPa, 117MPa, 137MPa and 157MPa at 400 ℃, the uniaxial tensile creep test time is longer than 200h, the pressures are respectively 117MPa, 137MPa and 157MPa at 350 ℃, the uniaxial tensile creep test time is longer than 200h, and the pressures are respectively 117MPa, 137MPa and 157MPa at 300 ℃, and the uniaxial tensile creep test time is longer than 200h.
2. The zirconium-based alloy for nuclear reactor fuel cladding having both corrosion and creep resistance according to claim 1, wherein the raw materials of the zirconium-based alloy comprise zirconium sponge for nuclear use, ferrovanadium for zirconium, niobium for zirconium and ferrous sulfide or elemental sulfur.
3. The zirconium based alloy for nuclear reactor fuel cladding having both corrosion and creep resistance according to claim 1 or 2, wherein the zirconium based alloy consists of the following components in mass percent: nb 0.2%, V0.05%, fe 0.5%, O0.16%, S0.002%, C70 ppm or less, si 70ppm or less, ti 40ppm or less, N35 ppm or less, and Zr and unavoidable impurities in balance.
4. The zirconium based alloy for nuclear reactor fuel cladding having both corrosion and creep resistance according to claim 1 or 2, wherein the zirconium based alloy consists of the following components in mass percent: nb is 2.0%, V is 0.11%, fe is 0.2%, O is 0.09%, S is 0.003%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
5. The zirconium based alloy for nuclear reactor fuel cladding having both corrosion and creep resistance according to claim 1 or 2, wherein the zirconium based alloy consists of the following components in mass percent: nb is 1.2%, V is 0.07%, fe is 0.15%, O is 0.14%, S is 0.0015%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
6. The zirconium based alloy for nuclear reactor fuel cladding having both corrosion and creep resistance according to claim 1 or 2, wherein the zirconium based alloy consists of the following components in mass percent: nb is 0.9%, V is 0.01%, fe is 0.03%, O is 0.16%, S is 0.0035%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
7. The zirconium based alloy for nuclear reactor fuel cladding having both corrosion and creep resistance according to claim 1 or 2, wherein the zirconium based alloy consists of the following components in mass percent: nb is 0.8%, V is 0.04%, fe is 0.12%, O is 0.08%, S is 0.03%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
8. The zirconium based alloy for nuclear reactor fuel cladding having both corrosion and creep resistance according to claim 1 or 2, wherein the zirconium based alloy consists of the following components in mass percent: nb is 1.1%, V is 0.005%, fe is 0.014%, O is 0.09%, S is 0.007%, C is not more than 70ppm, si is not more than 70ppm, ti is not more than 40ppm, N is not more than 35ppm, and the balance is Zr and unavoidable impurities.
9. A method for preparing a zirconium based alloy pipe, characterized in that the method for preparing the zirconium based alloy pipe selects the zirconium based alloy as defined in any one of claims 1 to 8 for preparation, and specifically comprises the following steps:
selecting raw materials according to the components of the zirconium-based alloy, and smelting the raw materials into a zirconium-iron-vanadium intermediate alloy and a zirconium-niobium intermediate alloy;
preparing an alloy bag from zirconium-iron-vanadium alloy and ferrous sulfide or elemental sulfur by using a zirconium foil, uniformly placing the alloy bag in the middle of zirconium sponge, pressing the alloy bag into an electrode, and carrying out conventional smelting for three times in a vacuum consumable arc furnace to obtain an alloy cast ingot;
step three, cogging and forging the alloy ingot obtained in the step two at the forging temperature of 960-1070 ℃ to obtain a bar blank;
and fourthly, carrying out solution treatment on the rod blank obtained in the third step at the temperature of 1000-1070 ℃, extruding the rod blank into a tube blank at the temperature of 600-700 ℃ after quenching, and finally obtaining the zirconium-based alloy tube after multi-pass cold rolling, intermediate annealing and finished product annealing.
10. The method for producing a zirconium based alloy pipe as claimed in claim 9, wherein the cold rolling passes in the fourth step are each 45% -85%, the intermediate annealing temperature is 560 ℃ -650 ℃, the final annealing temperature is 540 ℃ -600 ℃, the time is 2 h-5 h, and the vacuum degree is 8.0x10 or less -2 Under Pa.
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