CN113201665A - Zirconium alloy for fuel assembly cladding, manufacturing method thereof and fuel assembly cladding tube - Google Patents
Zirconium alloy for fuel assembly cladding, manufacturing method thereof and fuel assembly cladding tube Download PDFInfo
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- CN113201665A CN113201665A CN202110377319.6A CN202110377319A CN113201665A CN 113201665 A CN113201665 A CN 113201665A CN 202110377319 A CN202110377319 A CN 202110377319A CN 113201665 A CN113201665 A CN 113201665A
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- zirconium alloy
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- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22C—ALLOYS
- C22C16/00—Alloys based on zirconium
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- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22C—ALLOYS
- C22C1/00—Making non-ferrous alloys
- C22C1/02—Making non-ferrous alloys by melting
-
- 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
-
- 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
-
- 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
- G21C3/02—Fuel elements
- G21C3/04—Constructional details
- G21C3/06—Casings; Jackets
- G21C3/07—Casings; Jackets characterised by their material, e.g. alloys
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- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Abstract
The invention discloses a zirconium alloy for fuel assembly cladding, a manufacturing method thereof and a cladding tube of a fuel assembly, wherein the zirconium alloy for the fuel assembly cladding comprises the following components in percentage by mass: 0.45 to 0.95 percent of niobium, 0.21 to 0.35 percent of tin, 0.03 to 0.1 percent of iron, 0.03 to 0.1 percent of vanadium, 1000 to 1600ppm of oxygen and the balance of zirconium; wherein the total amount of iron and vanadium is less than or equal to 0.15 percent. The zirconium alloy has excellent corrosion resistance and good embrittlement resistance after high-temperature oxidation quenching through the proportion of each component, is suitable for a cladding material of a nuclear power station reactor, and improves the service performance and safety of a fuel assembly.
Description
Technical Field
The invention relates to the technical field of nuclear fuel, in particular to a zirconium alloy for fuel assembly cladding, a manufacturing method thereof and a cladding tube of a fuel assembly.
Background
The development of zirconium alloys for nuclear fuel assemblies has now been repeated for three generations of commercial zirconium alloys. The first generation was traditional Zr-4 and Zr-2, both of which were widely used in nuclear reactors since the 50 s of the last century. The second generation was low tin Zr-4 and optimized Zr-4. In 1977 to 1984, Siemens performed a significant number of cell-edge inspections of Zry-4 alloys and found that Sn, C, and Si contents had an effect on corrosion performance, with lower Sn and C contents being beneficial for corrosion performance and lower amounts of Si being beneficial. Siemens introduced a low tin Zr-4 alloy in 1986. Siemens later developed optimized Zr-4 based on low-tin Zr-4 alloys, which had higher Fe and Cr contents and better corrosion resistance than the low-tin Zr-4 alloys. The third generation is ZIRLO alloy developed by American West House company and M5 alloy developed by French Ashi enamel company, the corrosion resistance of the ZIRLO alloy is improved to a certain extent compared with that of the optimized Zr-4 alloy, the corrosion resistance of the M5 alloy is obviously improved compared with that of the optimized Zr-4 alloy, but the ZIRLO alloy has poor corrosion resistance in a high Li concentration environment and low adaptability to water chemistry, and the creep resistance of the M5 alloy is general.
The existing research shows that the performance of the existing zirconium alloy is not optimal at present, and the component proportion has a space for further optimization. The ZIRLO alloy is added with 1 wt% of Sn, so that the ZIRLO alloy has stronger creep resistance, and meanwhile, the corrosion resistance of the ZIRLO alloy has an obvious improvement space due to the high Sn content.
The corrosion resistance of the ZIRLO alloy is Improved to a certain extent after the Sn content is reduced (Yueh, H.K., Kesterson, R.L., Commtock, R.J., et al, Improved ZIRLOTM cladding performance chemistry and process modifications, zirconium in the Nuclear Industry, ASTM STP1467,2004, pp.330-346.). The Effect of Nb, Sn on the Oxidation of Zirconium Alloy Fuel Cladding in High Temperature water vapor is reported by a.malgin, russia in the 18 International Nuclear Industry Zirconium forum (abbreviated as "Zirconium association") (Malgin, a., Markelov, v.a., Gusev, a., Nikulina, a., Novikov, v.s., shelelov, i., Donnikov, v.s., latinin, v.and kosiina, j., "Alloying Effect of nickel and Tin on the Zirconium Alloy Cladding class laminate High Temperature Oxidation in stem," Alloying in the Nuclear reactor Industry index: 18th International Alloy, ASTM STP, r.j. copper Alloy transition Temperature, copper Alloy Temperature, copper Alloy, aluminum Alloy, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper, aluminum, copper.
In order to meet the increasing demand of fuel, the fuel cannot be optimally designed only in one performance aspect, and not only the conventional working condition but also the accident working condition need to be considered. With the development of zirconium alloy, the nuclear industry has higher and higher attention to the cladding behavior of zirconium alloy under accident conditions, particularly loss of coolant accident, particularly the embrittlement resistance after high-temperature oxidation quenching. Therefore, it is necessary to develop a zirconium alloy having excellent performance under various conditions by optimizing the composition ratio and adding other elements.
Disclosure of Invention
The invention aims to provide a zirconium alloy for fuel assembly cladding with excellent corrosion resistance and high-temperature oxidation quenching embrittlement resistance, a manufacturing method thereof and a fuel assembly cladding tube made of the zirconium alloy.
The technical scheme adopted by the invention for solving the technical problems is as follows: the zirconium alloy for the fuel assembly cladding comprises the following components in percentage by mass: 0.45 to 0.95 percent of niobium, 0.21 to 0.35 percent of tin, 0.03 to 0.1 percent of iron, 0.03 to 0.1 percent of vanadium, 1000 to 1600ppm of oxygen and the balance of zirconium; wherein the total amount of iron and vanadium is less than or equal to 0.15 percent.
Preferably, in the zirconium alloy, C is less than or equal to 100ppm, and N is less than or equal to 45 ppm.
The invention also provides a preparation method of the zirconium alloy, which comprises the following steps:
s1, providing raw materials respectively containing zirconium, niobium, tin, iron and vanadium, and weighing the raw materials according to the mass percentage of each component in the zirconium alloy;
s2, smelting the raw materials to obtain an alloy ingot;
s3, forging the alloy ingot into a blank;
s4, carrying out beta quenching on the blank;
s5, carrying out multiple cold rolling on the blank subjected to beta quenching, and carrying out intermediate annealing between each cold rolling;
and S6, performing final annealing on the cold-rolled blank to obtain the zirconium alloy.
Preferably, in step S3, the forging temperature is 800-1100 ℃.
Preferably, in step S4, the temperature of the β -quenching is 950 ℃ to 1100 ℃.
Preferably, in step S5, the temperature of the intermediate annealing is 550 ℃ to 600 ℃.
Preferably, the billet is extruded or hot rolled before being cold rolled in step S5.
Preferably, in step S6, the temperature of the final annealing is 460 ℃ to 600 ℃.
Preferably, the manufacturing method further comprises the following steps:
and S7, processing the zirconium alloy into a cladding tube.
The invention also provides a cladding tube of a fuel assembly, which is made of the zirconium alloy.
Compared with the existing Zr-4 alloy, the zirconium alloy has better corrosion resistance and good embrittlement resistance after high-temperature oxidation quenching through the proportion of the components, is suitable for a cladding material of a nuclear power station reactor, and improves the service performance and safety of a fuel assembly.
Detailed Description
The invention relates to a zirconium alloy for fuel assembly cladding, which is a low-tin zirconium tin niobium iron vanadium alloy and comprises the following components in percentage by mass:
0.45 to 0.95 percent of Nb (niobium), 0.21 to 0.35 percent of Sn (tin), 0.03 to 0.1 percent of Fe (iron), 0.03 to 0.1 percent of V (vanadium), 1000 to 1600ppm of O (oxygen), and the balance of Zr (zirconium). Wherein the total amount of iron and vanadium is less than or equal to 0.15 percent.
The zirconium alloy for fuel assembly cladding of the present invention further comprises: c (carbon) is less than or equal to 100ppm, and N (nitrogen) is less than or equal to 45 ppm. It will be appreciated that some unavoidable and small amounts of impurities are also included.
The zirconium alloy for the fuel assembly cladding is oxidized at 1204 ℃, and the residual plasticity value after quenching is more than 3.4 percent when the oxidation rate reaches 18 percent.
As for Nb (niobium), research shows that solid solution of niobium in the zirconium alloy is beneficial to the corrosion resistance and creep resistance of the zirconium alloy, but the content of niobium is sensitive to heat treatment when the content of niobium is too high, so that in the invention, in order to ensure that the zirconium alloy has excellent corrosion resistance and creep resistance, the content of Nb is controlled to be 0.45-0.95 wt%, and the full solid solution of niobium in the zirconium alloy is ensured.
Tin (Sn) has a relatively high solid solubility in zirconium, and when a certain amount of tin is added, the strength and creep resistance of the zirconium alloy are improved, but the addition of tin reduces the uniform corrosion resistance of the zirconium alloy. On the other hand, the addition of tin can improve the corrosion resistance of the zirconium alloy in a high Li concentration environment. The method comprehensively considers the influence of tin on the corrosion resistance and the corrosion resistance in a high Li environment when determining the tin content, and controls the tin content to be 0.21-0.35 wt%, thereby not only improving the adaptability of the zirconium alloy to water chemistry, but also reducing the adverse effect of tin on the corrosion resistance to the minimum extent, and leading the zirconium alloy to have excellent corrosion resistance.
Iron (Fe) and vanadium (V) are transition metal elements, and can increase the corrosion resistance of the zirconium alloy when added into the zirconium alloy, wherein the addition of the vanadium element can improve the hydrogen absorption resistance of the zirconium alloy. The transition metal elements of iron and vanadium need to be added in a proper amount in the zirconium alloy, and when the addition is excessive, the embrittlement resistance of the zirconium alloy is obviously reduced after high-temperature oxidation quenching. Therefore, in the invention, the contents of elements such as Fe, V and the like are strictly controlled, the content of Fe is controlled to be 0.03-0.1 wt%, the content of V is controlled to be 0.03-0.1 wt%, and the total amount of Fe and V in the zirconium alloy is less than or equal to 0.15 wt%, so that the residual plasticity value of the alloy after quenching is more than 3.4% when the alloy is oxidized at 1204 ℃ and the oxidation rate reaches 18% (calculated by using a Cathcart-Pawel formula (zirconium alloy oxidation law)), and the zirconium alloy has enough embrittlement resistance after high-temperature oxidation quenching.
In the zirconium alloy of the present invention, the addition of oxygen (O) can improve the strength and creep resistance of the zirconium alloy, but as the oxygen content increases, the workability, particularly the punching resistance, of the zirconium alloy decreases. Therefore, the oxygen content is controlled to 1000ppm to 1600 ppm.
The preparation method of the zirconium alloy can comprise the following steps:
and S1, providing raw materials respectively containing zirconium, niobium, tin, iron and vanadium, and weighing the raw materials according to the mass percentage of each component in the zirconium alloy (calculating by batching).
For example, nuclear grade sponge zirconium is used as the zirconium raw material. The elements niobium, tin, iron and vanadium are added in the form of pure metals or master alloys. And S2, smelting the raw materials to obtain an alloy ingot.
Putting all the raw materials into a vacuum smelting furnace for smelting, adjusting the content of O, C and N, and finally preparing an alloy ingot.
S3, forging the alloy ingot into a blank at the temperature of 800-1100 ℃.
And S4, carrying out beta quenching on the blank.
Wherein the temperature of beta quenching is 950-1100 ℃, and the temperature is kept for a long enough time to ensure that the whole blank reaches the quenching temperature. And S5, carrying out cold rolling on the blank subjected to beta quenching for multiple times, and carrying out intermediate annealing between cold rolling for each time.
Wherein, before cold rolling, the blank is extruded or hot rolled according to the shape of the zirconium alloy (such as a pipe material and the like) to be formed, and then the blank is subjected to cold rolling for at least 4 times. The temperature of the intermediate annealing is 550-600 ℃.
S6, carrying out final annealing on the cold-rolled blank at 460-600 ℃ to obtain the zirconium alloy.
The zirconium alloy can be made into profiles, plates or pipes according to the requirements of application products.
For example, the manufacturing method of the present invention further includes the steps of:
s7, processing the zirconium alloy prepared in the step S6 into a cladding tube for a fuel assembly.
In one embodiment, the zirconium alloy described above is formed into a cladding tube for a fuel assembly.
The present invention is further illustrated by the following specific examples.
The zirconium alloys of examples 1 to 4 were prepared by the manufacturing method of the present invention, and the contents of the respective components in the zirconium alloys of examples 1 to 4 are shown in table 1.
TABLE 1
The zirconium alloys and Zr-4 alloys (Zr-1.30Sn-0.20Fe-0.10Cr-0.12O) obtained in examples 1 to 4 and comparative examples 1 to 2 were subjected to corrosion test as comparative example 3. The corrosion test was carried out on an autoclave under corrosion conditions of 360 ℃/18.6 MPa/deionized water for a test period of 130 days. The results are shown in table 2 below.
TABLE 2
| Examples | Amount of corrosion (mg/dm)2) |
| 1 | 44.62 |
| 2 | 49.15 |
| 3 | 50.21 |
| 4 | 46.12 |
| Comparative example 1 | 45.33 |
| Comparative example 2 | 45.96 |
| Comparative example 3 | 63.30 |
As is apparent from the data shown in Table 2, the zirconium alloys of examples 1 to 4 and comparative examples 1 to 2 have higher corrosion resistance than the conventional Zr-4 alloy.
The zirconium alloys obtained in examples 1 to 4 and comparative examples 1 to 2 and the Zr-4 alloy (Zr-1.30Sn-0.20Fe-0.10Cr-0.12O) as comparative example 3 were subjected to a ring crush test after oxidation quenching to observe the embrittlement resistance after high-temperature oxidation quenching (embrittlement resistance under LOCA accident conditions). The oxidation quenching process comprises the following steps: the test temperature is 1204 ℃, the temperature is maintained for a certain time, when the CP-ECR (equivalent zirconium reaction amount calculated by a Cathcart-Pawel formula) of the sample reaches 18%, the sample is slowly cooled to 800 ℃ within 200s, then quenching is carried out, a hoop compression test is carried out on the sample after oxidation quenching, and when the compensation strain value calculated according to a load displacement curve is more than 3.4%, the sample has enough plasticity. The results are shown in table 3 below.
TABLE 3
| Examples | Compensating strain (%) |
| 1 | 8.1 |
| 2 | 9.6 |
| 3 | 6.3 |
| 4 | 5.5 |
| Comparative example 1 | 3.1 |
| Comparative example 2 | 3.4 |
| Comparative example 3 | 4.0 |
The compensation strain value obtained by the hoop compression test after oxidation quenching reflects the residual plasticity (namely the embrittlement resistance after high-temperature oxidation quenching) of the test material after quenching, and the data shown in Table 3 show that the total content of iron and vanadium in examples 3 and 4 is higher than that of iron and vanadium in examples 1 and 2, and the compensation strain is obviously lower, so the embrittlement resistance after high-temperature oxidation quenching is not as good as that of examples 1 and 2, but is better than that of Zr-4 alloy in comparative example 3, and has excellent embrittlement resistance after high-temperature oxidation quenching compared with that of Zr-4 alloy. In comparative examples 1 and 2, since the total content of iron and vanadium exceeded 0.15 wt%, the embrittlement resistance after high-temperature oxidation quenching was inferior to that of examples 1 to 4 and comparative example 3. The total content of iron and vanadium in the zirconium alloys obtained in the examples 1 to 4 is not more than 0.15 wt%, and the compensation strain values are all more than 3.4%, which shows that the zirconium alloys obtained in the examples 1 to 4 have excellent embrittlement resistance after high-temperature oxidation quenching.
It can be understood that, in addition to the above embodiments, the zirconium alloy of the present invention, which has excellent corrosion resistance and embrittlement resistance after high-temperature oxidation quenching within the range of the content of each component and satisfies the content relation, is suitable for being used as a nuclear power plant reactor cladding material.
The above description is only an embodiment of the present invention, and not intended to limit the scope of the present invention, and all modifications of equivalent structures and equivalent processes, which are made by the present specification, or directly or indirectly applied to other related technical fields, are included in the scope of the present invention.
Claims (10)
1. The zirconium alloy for the fuel assembly cladding is characterized by comprising the following components in percentage by mass: 0.45 to 0.95 percent of niobium, 0.21 to 0.35 percent of tin, 0.03 to 0.1 percent of iron, 0.03 to 0.1 percent of vanadium, 1000 to 1600ppm of oxygen and the balance of zirconium; wherein the total amount of iron and vanadium is less than or equal to 0.15 percent.
2. The zirconium alloy for fuel assembly cladding as set forth in claim 1 wherein C is 100ppm or less and N is 45ppm or less.
3. A method of making a zirconium alloy for fuel assembly cladding as claimed in claim 1 or claim 2, comprising the steps of:
s1, providing raw materials respectively containing zirconium, niobium, tin, iron and vanadium, and weighing the raw materials according to the mass percentage of each component in the zirconium alloy;
s2, smelting the raw materials to obtain an alloy ingot;
s3, forging the alloy ingot into a blank;
s4, carrying out beta quenching on the blank;
s5, carrying out multiple cold rolling on the blank subjected to beta quenching, and carrying out intermediate annealing between each cold rolling;
and S6, performing final annealing on the cold-rolled blank to obtain the zirconium alloy.
4. The method of making a zirconium alloy for fuel assembly cladding as set forth in claim 3, wherein said forging is at a temperature of 800 ℃ to 1100 ℃ in step S3.
5. The method of making a zirconium alloy for fuel assembly cladding as set forth in claim 3, wherein said β -quenching is at a temperature of 950 ℃ to 1100 ℃ in step S4.
6. The method of making a zirconium alloy for fuel assembly cladding as set forth in claim 3, wherein said intermediate annealing is performed at a temperature of 550 ℃ to 600 ℃ in step S5.
7. The method of claim 3, wherein the billet is extruded or hot rolled prior to cold rolling in step S5.
8. The method of making a zirconium alloy for fuel assembly cladding as set forth in claim 3, wherein said final annealing is at a temperature of 460 ℃ to 600 ℃ in step S6.
9. The method of making a zirconium alloy for fuel assembly cladding as set forth in any one of claims 3 to 8, further including the steps of:
and S7, processing the zirconium alloy into a cladding tube.
10. A cladding tube for a fuel assembly, characterized by being made of the zirconium alloy as set forth in claim 1 or 2.
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| CN202110377319.6A CN113201665A (en) | 2021-04-08 | 2021-04-08 | Zirconium alloy for fuel assembly cladding, manufacturing method thereof and fuel assembly cladding tube |
| PCT/CN2021/117834 WO2022213544A1 (en) | 2021-04-08 | 2021-09-10 | Zirconium alloy and fabrication method therefor, cladding tube, and fuel assembly |
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| CN202110377319.6A CN113201665A (en) | 2021-04-08 | 2021-04-08 | Zirconium alloy for fuel assembly cladding, manufacturing method thereof and fuel assembly cladding tube |
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| CN115652237A (en) * | 2022-08-16 | 2023-01-31 | 重庆大学 | Zirconium alloy containing triple twin crystals and preparation method thereof |
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| CN104745875A (en) * | 2013-12-30 | 2015-07-01 | 上海核工程研究设计院 | Zirconium alloy material for light water reactor under higher burnup |
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| FR2624136B1 (en) * | 1987-12-07 | 1992-06-05 | Cezus Co Europ Zirconium | TUBE, BAR OR SHEET IN ZIRCONIUM ALLOY, RESISTANT TO BOTH UNIFORM CORROSION AND NODULAR CORROSION AND METHOD OF MANUFACTURE THEREOF |
| DE10332239B3 (en) * | 2003-07-16 | 2005-03-03 | Framatome Anp Gmbh | Zirconium alloy and components for the core of light water cooled nuclear reactors |
| CN101270425B (en) * | 2008-03-24 | 2010-08-11 | 中国核动力研究设计院 | Zirconium based alloy for light-water reactor |
| CN103898366B (en) * | 2012-12-27 | 2016-08-10 | 中国核动力研究设计院 | A kind of zirconium-base alloy for power producer fuel assembly |
| CN105441717B (en) * | 2016-01-06 | 2017-09-26 | 中国核动力研究设计院 | A kind of nuclear power core structural material zirconium-base alloy |
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| CN104745875A (en) * | 2013-12-30 | 2015-07-01 | 上海核工程研究设计院 | Zirconium alloy material for light water reactor under higher burnup |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN115652237A (en) * | 2022-08-16 | 2023-01-31 | 重庆大学 | Zirconium alloy containing triple twin crystals and preparation method thereof |
| CN115652237B (en) * | 2022-08-16 | 2023-11-24 | 重庆大学 | Zirconium alloy containing triple twin crystals and preparation method thereof |
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