CN105568056B - Zirconium alloy for pressurized water reactor fuel element cladding - Google Patents
Zirconium alloy for pressurized water reactor fuel element cladding 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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- G—PHYSICS
- 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
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Abstract
The invention discloses a zirconium alloy for a pressurized water reactor fuel element cladding, which comprises the following components in percentage by weight: sn: 0.2-0.5%, Nb: 0.4-0.8%, Fe: 0.1-0.5%, Cr: 0.15-0.35%, V or Cu or Ni: 0.01-0.2%, Mo or S: 0.01-0.1%, O: 0.06-0.15%, C: less than 0.008%, N: less than 0.006%, and the balance Zr and other impurities. On the basis of Zr-Sn-Nb alloy, other element components for improving the alloy performance are added, and proper component content is selected, the alloy performance provided by the invention meets the requirement of high fuel consumption of a nuclear power reactor on a reactor core structure material, and the corrosion resistance of a product prepared from the prototype alloy in pure water outside the reactor and a boron-containing lithium-containing aqueous solution is improved.
Description
Technical Field
The invention belongs to the technical field of special alloy materials, and particularly relates to a zirconium alloy material for a pressurized water reactor fuel element cladding.
Background
The zirconium alloy has small thermal neutron absorption cross section, good corrosion resistance in high-temperature and high-pressure water and steam and good neutron irradiation resistance in a reactor, so that the zirconium alloy is commonly used as a cladding material of a nuclear power water-cooled reactor and is also the only cladding material adopted by the reactor of a nuclear power station at present. In the development of light water reactors, fuel design has placed high demands on reactor core structural components, such as fuel element cladding, grids, guide tubes, etc. Early, cladding materials were typically made from Zr-4 alloys, and later high fuel burn designs required increased coolant temperatures and extended residence time of the zirconium alloy cladding in the stack, thereby exposing the zirconium alloy cladding to more severe corrosive environments, these high requirements prompted research into improving corrosion resistance of Zr-4 alloys, while also driving the development of new zirconium alloys with superior corrosion resistance.
With the further development of nuclear power, on the basis of ensuring the safety of the nuclear reactor, the economic efficiency of the nuclear reactor needs to be improved, and the nuclear power operation cost needs to be reduced, so that the aims of long service life, high fuel consumption and zero damage are provided for fuel assemblies. This means that the zirconium alloy cladding has increased water side corrosion, increased hydrogen absorption, increased irradiation time, increased interaction between pellets and cladding, increased internal pressure, etc., which puts higher demands on the performance of the zirconium alloy. Aiming at the high requirement of the nuclear power technology development on the fuel cladding, the research on novel zirconium alloy is internationally developed, and novel zirconium alloys such as ZIRLO, E635, M5, X5A and the like with better corrosion resistance than Zr-4 alloy are obtained. The existing research shows that the proportion of the components in the existing zirconium alloy is not necessarily in the optimal range, for example, after the Sn content in the ZIRLO alloy is reduced, the corrosion resistance of the ZIRLO alloy is further improved; the HANA-6 alloy formed by adding a trace amount of Cu (0.05 wt%) into the Zr-Nb alloy also has very good corrosion resistance; abnormal phenomena such as bending of fuel rods or fuel assemblies, poor radiation growth resistance and the like occur in the M5 alloy in the in-pile operation process, so that in France, a small amount of Sn and Fe are added on the basis of the M5 alloy components, and the mechanical properties, particularly the creep and radiation growth properties, of the alloy are greatly improved on the basis of keeping the excellent corrosion resistance of the alloy. Therefore, the zirconium alloy with better corrosion resistance can be developed by optimizing the mixture ratio of alloy components or adding other alloy elements on the basis of the existing zirconium alloy so as to meet the requirement of continuously improving the fuel consumption.
In addition, after the components of the alloy are determined, the corrosion resistance of the alloy can be further improved by adopting a proper hot working process. For example, in zirconium alloys with high Nb content, including ZIRLO, M5, N36, etc., when the hot working temperature is raised, the corrosion resistance is deteriorated due to coarsening and uneven distribution of the second phase and supersaturated solid solution of Nb in the alloy matrix, so that the 'low temperature processing technology', i.e., the low temperature processing technology with lower hot working temperature and annealing temperature can obtain fine and dispersed second phase structures, is emphasized, thereby greatly improving the corrosion resistance and mechanical properties of the alloy.
Disclosure of Invention
The invention further optimizes the components and the proportion of the prior zirconium alloy to obtain a novel zirconium alloy with good corrosion resistance.
In order to realize the purpose, the invention adopts the technical scheme that:
the zirconium alloy for the cladding of the pressurized water reactor fuel element comprises the following components in percentage by weight: sn: 0.2-0.5%, Nb: 0.4-0.8%, Fe: 0.1-0.5%, Cr: 0.15-0.35%, V or Cu or Ni: 0.01-0.2%, Mo or S: 0.01-0.1%, O: 0.06-0.15%, C: less than 0.008%, N: less than 0.006%, and the balance Zr and other impurities.
The zirconium alloy for the cladding of the pressurized water reactor fuel element comprises the following components in percentage by weight: sn: 0.2-0.5%, Nb: 0.4-0.8%, Fe: 0.1-0.5%, Cr: 0.15-0.35%, V or Cu or Ni: 0.01-0.2%, O: 0.06-0.15%, C: less than 0.008%, N: less than 0.006%, and the balance Zr and other impurities.
The zirconium alloy for the cladding of the pressurized water reactor fuel element comprises the following components in percentage by weight: sn: 0.2-0.5%, Nb: 0.4-0.8%, Fe: 0.1-0.5%, Cr: 0.15-0.35%, Mo or S: 0.01-0.1%, O: 0.06-0.15%, C: less than 0.008%, N: less than 0.006%, and the balance Zr and other impurities.
The zirconium alloy for the cladding of the pressurized water reactor fuel element comprises the following components in percentage by weight: sn: 0.15-0.35%, Nb: 0.4-0.6%, Fe: 0.2-0.4%, Cr: 0.05-0.25%, V or Cu or Ni: 0.01-0.2%, Mo or S: 0.01-0.1%, O: 0.06-0.15%, C: less than 0.008%, N: less than 0.006%, and the balance Zr and other impurities.
The zirconium alloy for the cladding of the pressurized water reactor fuel element comprises the following components in percentage by weight: sn: 0.1-0.3%, Nb: 0.6-0.8%, Fe: 0.05-0.25%, Cr: 0.05-0.15%, V or Cu or Ni: 0.01-0.2%, Mo or S: 0.01-0.1%, O: 0.06-0.15%, C: less than 0.008%, N: less than 0.006%, and the balance Zr and other impurities.
The zirconium alloy for the cladding of the pressurized water reactor fuel element comprises the following components in percentage by weight: sn: 0.25%, Nb: 0.5%, Fe: 0.3%, Cr: 0.12%, O: 0.12%, C: less than 0.008%, N: less than 0.006%, and the balance Zr and other impurities.
The zirconium alloy for the cladding of the pressurized water reactor fuel element comprises the following components in percentage by weight: sn: 0.2%, Nb: 0.7%, Fe: 0.15%, Cr: 0.1%, V: 0.15%, O: 0.12%, C: less than 0.008%, N: less than 0.006%, and the balance Zr and other impurities.
The invention takes Zr-Sn-Nb alloy system as a base, and other alloy elements such as Fe, Cr, V or Cu or Ni, Mo or S are added into the alloy system in a multi-element and small-amount mode. In the design of the alloy, the alloy elements are reasonably combined and matched with each other, unexpected effects can be generated on the corrosion resistance of the alloy, the corrosion resistance outside the pile of the alloy can be greatly improved, and the alloy is expected to have excellent corrosion resistance, irradiation growth resistance and creep deformation resistance in the pile.
The preparation method of the zirconium alloy for the pressurized water reactor fuel element cladding comprises the following steps:
(1) preparing the components in the zirconium alloy according to the design components;
(2) smelting in a vacuum consumable arc furnace to prepare an alloy ingot;
(3) forging the alloy ingot into a blank material with a required shape in a beta phase region at 900-1050 ℃;
(4) heating and homogenizing the blank in a beta phase region at 1000-1100 ℃, and quenching;
(5) carrying out hot working on the quenched blank in a beta phase region at 600-700 ℃;
(6) performing cold processing on the blank after the hot processing, and performing intermediate annealing at the temperature of 560-650 ℃;
(7) and carrying out stress relief annealing or recrystallization annealing treatment at 480-620 ℃ to obtain the zirconium alloy material.
In conclusion, the beneficial effects of the invention are as follows: on the basis of Zr-Sn-Nb alloy, other element components for improving the alloy performance are added, the proper component content is selected, and the solid solution, phase component, second phase crystal structure, components and types are controlled, so that the alloy performance provided by the invention meets the requirement of high fuel consumption of a nuclear power reactor on a reactor core structure material, and the corrosion resistance of a product prepared from the prototype alloy in pure water outside the reactor and a boron-containing lithium-containing aqueous solution is improved. Through the test results in the specific embodiment, the alloy has more excellent corrosion resistance, higher creep and fatigue resistance and radiation growth resistance when used in a reactor.
Detailed Description
The present invention will be described in further detail with reference to examples, but the embodiments of the present invention are not limited thereto.
For zirconium alloys used as nuclear reactor cladding materials, the corrosion resistance of the zirconium alloy is a primary consideration, based on which the cost of production and the workability are considerations in selecting the alloying elements. Therefore, the impact of each alloying element on corrosion resistance, mechanical properties and creep behavior and the range of the amount of each alloying element in the alloy system need to be studied in detail. The zirconium alloy has better uniform and nodular corrosion resistance, higher creep resistance, fatigue resistance and radiation growth resistance.
The functions and the dosage of each alloy element are as follows:
(1) zirconium (Zr)
In view of the neutron economy, the invention selects zirconium with a small neutron absorption cross section (0.185 b) as the matrix element, and also considers the neutron absorption cross section of other alloy elements added into the zirconium matrix.
(2) Tin (Sn)
Tin stabilizes the alpha phase of zirconium, increases its strength, and counteracts the detrimental effects of nitrogen on corrosion. When the amount of tin used is too small, the desired effect cannot be achieved. The Sn content in the alloy is 0.2-0.5% (weight percentage), so that the alloy has excellent corrosion resistance and good mechanical property.
(3) Niobium (Nb)
Niobium can stabilize beta phase of zirconium and has higher strengthening effect on zirconium. When the amount of niobium is too large, it is sensitive to heat treatment. The Nb content of the invention is 0.4-0.8% (weight percentage), which can ensure the alloy to have excellent corrosion resistance and good mechanical property.
(4) Iron (Fe)
Iron improves the corrosion resistance and mechanical properties of the alloy, but too much or too little iron can have an adverse effect. The content of Fe added in the invention is 0.1-0.5% (weight percentage), which can ensure that the alloy has excellent corrosion resistance.
(5) Chromium (Cr)
Chromium improves the corrosion resistance and mechanical properties of the alloy, but too much or too little chromium can have an adverse effect. The content of the Cr added in the invention is 0.15-0.35% (weight percentage), which can ensure that the alloy has excellent corrosion resistance.
(6) Vanadium (V)
Vanadium can improve the corrosion resistance and mechanical property of the alloy, and the corrosion property of the vanadium-zirconium-containing alloy is sensitive to heat treatment. The content of V in the invention is 0.01-0.2% (weight percentage), which can ensure the alloy has excellent corrosion resistance.
(7) Copper (Cu)
Copper can improve the corrosion resistance of the alloy, but excessive amounts of copper can have an adverse effect. The Cu content of the alloy is 0.01-0.2% (weight percentage), and the alloy can be ensured to have excellent corrosion resistance.
(8) Nickel (Ni)
Nickel can improve the corrosion resistance of the alloy, but too much nickel can have an adverse effect. The content of the Ni added in the invention is 0.01-0.2% (weight percentage), which can ensure that the alloy has excellent corrosion resistance.
(9) Molybdenum (Mo)
Molybdenum has a high strengthening effect on zirconium, but the plasticity is reduced, and a ZrMo2 second phase generated by Mo and Zr is intensively distributed in a matrix, so that the corrosion resistance is not good, but the dispersion strengthening effect on zirconium alloy is good. The content of Mo added in the invention is 0.01-0.1% (weight percentage), which can ensure that the alloy has excellent corrosion resistance.
(10) Sulfur (S)
The addition of a proper amount of sulfur in the alloy can improve the creep strength of the alloy and improve the corrosion resistance of the alloy. However, too much sulfur can have adverse effects. In the invention, the content of S added is 0.01-0.1% (weight percentage), which can ensure that the alloy has excellent corrosion resistance.
(11) Oxygen (O)
Oxygen can stabilize the alpha phase of zirconium, and the yield strength can be improved by adding oxygen into the alloy. The content of O added in the invention is 0.06-0.15% (weight percentage), which can ensure that the alloy has enough mechanical property and creep resistance. The control difficulty in the material processing process is greatly reduced due to the increase of the oxygen content.
(12) Carbon (C)
Carbon in the alloy is present as an inevitable impurity element, and when the content is high, the corrosion resistance of the alloy is lowered. The weight percentage of C in the invention is less than 0.008 percent, which can ensure that the alloy has excellent corrosion resistance in high-temperature water and steam.
(13) Nitrogen (N)
Nitrogen in the alloy is present as an inevitable impurity element, and when the content is high, the corrosion resistance of the alloy is lowered. The weight percentage of N in the invention is less than 0.006 percent, which can ensure that the alloy has excellent corrosion resistance in high-temperature water and steam.
Table 1 shows the composition (in weight%) of the alloys provided by the present invention, 9 and 10 in the table are the composition of Zr-4 alloy and N36 alloy, respectively, and the balance of each alloy provided in table 1 is Zr and other impurities.
The alloy provided in table 1 above was prepared as follows:
(1) preparing the components in the zirconium alloy according to the design components;
(2) smelting in a vacuum consumable arc furnace to prepare an alloy ingot;
(3) forging the alloy ingot into a blank material with a required shape in a beta phase region at 900-1050 ℃;
(4) heating and homogenizing the blank in a beta phase region at 1000-1100 ℃, and quenching;
(5) carrying out hot working on the quenched blank in a beta phase region at 600-700 ℃;
(6) performing cold processing on the blank after the hot processing, and performing intermediate annealing at the temperature of 560-650 ℃;
(7) and carrying out stress relief annealing or recrystallization annealing treatment at 480-620 ℃ to obtain the zirconium alloy material.
The zirconium alloy material prepared by the processing technology has a microstructure consisting of equiaxial alpha-Zr crystal grains and uniformly distributed fine second phase particles, and can ensure that the zirconium alloy material has excellent service performance in a harsh environment of a reactor core. The results of testing the corrosion performance of the zirconium alloy material prepared by the method are shown in table 2, and are respectively the corrosion rate of the alloy material provided by the invention after being corroded in pure water at 360 ℃ for 300 days and the corrosion rate of the alloy material after being corroded in lithium-containing water at 360 ℃ for 300 days.
As can be seen from Table 2, the alloy material provided by the invention shows good corrosion resistance in pure water and boron-containing lithium-containing aqueous solution at 360 ℃.
In conclusion, the application examples provided by the invention show that the alloy of the invention shows good corrosion resistance when corroded under the 2 water chemistry conditions, and the alloy is pure water at 360 ℃/18.6MPa and LiOH + H at 360 ℃/18.6MPa3BO3The corrosion rate in aqueous solution is significantly better than that of the Zr-4 and N36 alloys.
Because the invention adopts the preferable composition ranges of Sn, Nb, Fe, Cr, V, Cu, Ni, Mo and S, the interaction between the alloy elements in the ranges and the low-temperature processing technology generate the effect which is not expected in advance, and the effect is mainly shown in two aspects: 1) the alloy of the invention shows good corrosion resistance when being corroded under the 2 kinds of water chemistry conditions, and is obviously superior to Zr-4 and N36 alloys; 2) the alloy of the invention obtains a fine dispersed second phase after being processed by a low-temperature process, and improves the mechanical properties (such as creep and fatigue properties) and the anti-radiation growth performance of the alloy.
As described above, the present invention can be preferably realized.
The foregoing is only a preferred embodiment of the present invention, and the present invention is not limited thereto in any way, and any simple modification, equivalent replacement and improvement made to the above embodiment within the spirit and principle of the present invention still fall within the protection scope of the present invention.
Claims (1)
1. The zirconium alloy for the cladding of the pressurized water reactor fuel element is characterized by comprising the following components in percentage by weight: sn: 0.5%, Nb: 0.63-0.8%, Fe: 0.41-0.5%, Cr: 0.33-0.35%, V or Cu or Ni: 0.01-0.02%, Mo or S: 0.01-0.1%, O: 0.06-0.15%, C: less than 0.008%, N: less than 0.006%, the balance being Zr and other impurities; corroding for 300 days in a pure water environment with the temperature of 360 ℃ and the pressure of 18.6MPa, wherein the corrosion rate of the zirconium alloy is 0.14 mg/(dm)2·d)。
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| CN109022915A (en) * | 2018-10-11 | 2018-12-18 | 上海核工程研究设计院有限公司 | A kind of high-performance zirconium-base alloy and preparation method thereof containing molybdenum element |
| CN110284027B (en) * | 2019-08-06 | 2020-04-21 | 中国核动力研究设计院 | Zirconium-based alloy resistant to alkaline water corrosion |
| CN112481521B (en) * | 2020-04-13 | 2021-08-31 | 国核宝钛锆业股份公司 | High-strength zirconium alloy and preparation method of bar for high-strength zirconium alloy fastener |
| CN115747570A (en) * | 2022-10-31 | 2023-03-07 | 上海大学 | Zirconium alloy cladding material for small pressurized water reactor and preparation method thereof |
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