WO2022213544A1 - 锆合金及其制作方法、包壳管及燃料组件 - Google Patents

锆合金及其制作方法、包壳管及燃料组件 Download PDF

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WO2022213544A1
WO2022213544A1 PCT/CN2021/117834 CN2021117834W WO2022213544A1 WO 2022213544 A1 WO2022213544 A1 WO 2022213544A1 CN 2021117834 W CN2021117834 W CN 2021117834W WO 2022213544 A1 WO2022213544 A1 WO 2022213544A1
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zirconium alloy
zirconium
vanadium
quenching
cladding tube
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PCT/CN2021/117834
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English (en)
French (fr)
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高长源
徐杨
陈刘涛
石林
陈敏莉
张利斌
王旭
邹红
聂立红
邓勇军
陈建新
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中广核研究院有限公司
岭澳核电有限公司
中国广核集团有限公司
中国广核电力股份有限公司
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Publication of WO2022213544A1 publication Critical patent/WO2022213544A1/zh

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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
    • 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
    • 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

Definitions

  • the invention relates to the technical field of nuclear fuels, in particular to a zirconium alloy and a manufacturing method thereof, a cladding tube and a fuel assembly.
  • zirconium alloys for nuclear fuel assemblies has now iterated three generations of commercial zirconium alloys.
  • the currently used zirconium alloys have strong creep resistance, but are also insufficient in corrosion resistance.
  • the technical problem to be solved by the present invention is to provide a zirconium alloy with excellent corrosion resistance and resistance to embrittlement after high-temperature oxidation quenching, a manufacturing method thereof, a cladding tube made of the zirconium alloy, and a cladding tube having the zirconium alloy. fuel assembly.
  • the technical solution adopted by the present invention to solve the technical problem is to provide a zirconium alloy, which includes the following components by mass percentage: niobium 0.45%-0.95%, tin 0.21%-0.35%, iron 0.03%-0.1%, vanadium 0.03% ⁇ 0.1%, also including oxygen 1000ppm-1600ppm, and the balance is zirconium; wherein, the total amount of iron and vanadium is ⁇ 0.15%.
  • the present invention also provides a method for making the above-mentioned zirconium alloy, comprising the following steps:
  • the temperature of the forging is 800°C-1100°C.
  • the temperature of the beta quenching is 950°C-1100°C.
  • the temperature of the intermediate annealing is 550°C-600°C.
  • step S5 the billet is extruded or hot rolled before cold rolling.
  • the temperature of the final annealing is 460°C-600°C.
  • the manufacturing method further comprises the following steps:
  • the present invention also provides a cladding tube, which is made of the zirconium alloy described in any one of the above.
  • the present invention also provides a fuel assembly comprising the cladding tube described above.
  • the zirconium alloy of the present invention has better corrosion resistance and better anti-embrittlement performance after high temperature oxidation quenching through the ratio of each component, and is suitable for the cladding material of nuclear power plant reactors. Service performance and safety of fuel assemblies.
  • the zirconium alloy of the present invention is a low tin zirconium tin niobium iron vanadium alloy, which comprises the following components by mass percentage:
  • the zirconium alloy of the present invention also includes: C (carbon) ⁇ 100ppm, N (nitrogen) ⁇ 45ppm. Understandably, some inevitable and small amounts of impurities are also included.
  • the residual plasticity value after quenching is greater than 3.4%.
  • Nb niobium
  • studies have shown that the solid solution niobium in the zirconium alloy is beneficial to the corrosion resistance and creep resistance of the zirconium alloy, but the high content of niobium will be sensitive to heat treatment. Therefore, in the present invention, in order to ensure the zirconium alloy The alloy has excellent corrosion resistance and creep resistance, and the content of Nb is controlled at 0.45wt% to 0.95wt%, so as to ensure the sufficient solid solution of niobium in the zirconium alloy.
  • Tin (Sn) has a large solid solubility in zirconium. After a certain amount of tin is incorporated, the strength and creep resistance of zirconium alloys will be improved, but the addition of tin will reduce the uniform corrosion resistance of zirconium alloys. On the other hand, the addition of tin can improve the corrosion resistance of zirconium alloys in high Li concentration environments.
  • the influence of tin on corrosion resistance and corrosion resistance in high Li environment is comprehensively considered when determining the tin content, and the tin content is controlled at 0.21wt% to 0.35wt%, which not only improves the adaptability of the zirconium alloy to water chemistry It also minimizes the adverse effects of tin on corrosion resistance, so that zirconium alloys have excellent corrosion resistance.
  • Iron (Fe) and vanadium (V) are transition metal elements, which can be added to the zirconium alloy to increase the corrosion resistance of the zirconium alloy.
  • the addition of vanadium can improve the hydrogen absorption resistance of the zirconium alloy.
  • Such transition group metal elements of iron and vanadium need to be added in an appropriate amount in zirconium alloys. If they are added too much, the embrittlement resistance of zirconium alloys after high temperature oxidation quenching will be significantly reduced.
  • the content of Fe, V and other elements is strictly controlled, the Fe content is controlled at 0.03wt% to 0.1wt%, the V content is controlled at 0.03wt% to 0.1wt%, and the content of iron and vanadium in the zirconium alloy is controlled at 0.03wt% to 0.1wt%.
  • the total amount is less than or equal to 0.15wt%, to ensure that the alloy is oxidized at 1204 °C and the oxidation rate reaches 18% (calculated using the Cathcart-Pawel formula (Zirconium Alloy Oxidation Law)) after quenching
  • the residual plasticity value is greater than 3.4%, ensuring that the zirconium alloy has Sufficient high temperature oxidation quenching resistance to embrittlement.
  • the addition of oxygen (O) can improve the strength and creep resistance of the zirconium alloy, but as the oxygen content increases, the machinability of the zirconium alloy will decrease, especially the punching resistance. Therefore, the oxygen content is controlled at 1000ppm-1600ppm.
  • the manufacturing method of the zirconium alloy of the present invention may comprise the following steps:
  • nuclear grade sponge zirconium is used as the zirconium raw material.
  • Niobium, tin, iron and vanadium elements are added in the form of pure metals or master alloys.
  • the alloy ingot is forged into a billet at a temperature of 800°C-1100°C.
  • the temperature of ⁇ quenching is 950°C-1100°C, and the temperature is kept for a long enough time to make the whole billet reach the quenching temperature.
  • S5. The billet after ⁇ quenching is subjected to multiple cold rolling, and intermediate annealing is carried out between each cold rolling.
  • the billet is extruded or hot rolled, and then the billet is subjected to at least 4 passes of cold rolling.
  • the temperature of the intermediate annealing is 550°C-600°C.
  • Zirconium alloys can be made into profiles, sheets or tubes according to the needs of the application.
  • the manufacturing method of the present invention also comprises the following steps:
  • step S7 The zirconium alloy obtained in step S6 is processed into a cladding tube for use in a fuel assembly.
  • the above-mentioned zirconium alloy is made into a cladding tube of a fuel assembly.
  • the fuel assembly includes the above-mentioned cladding tube made of zirconium alloy, and also includes fuel pellets sealed in the cladding tube. Since the cladding tube is made of the above-mentioned zirconium alloy, the cladding tube made of the conventional Zr-4 alloy has better corrosion resistance and better embrittlement resistance after high temperature oxidation quenching, thereby improving the service performance of the fuel assembly. safety.
  • the zirconium alloys of Examples 1 to 4 were prepared, and the contents of each component in the zirconium alloys of Examples 1 to 4 are shown in Table 1.
  • the zirconium alloys and Zr-4 alloys (Zr-1.30Sn-0.20Fe-0.10Cr-0.12O) prepared in Examples 1-4 and Comparative Examples 1-2 were used as Comparative Example 3 to conduct corrosion tests.
  • the corrosion test was carried out in an autoclave, the corrosion conditions were 360°C/18.6MPa/deionized water, and the test time was 130 days. The results are shown in Table 2 below.
  • the zirconium alloys prepared in Examples 1-4 and Comparative Examples 1-2 and the Zr-4 alloy (Zr-1.30Sn-0.20Fe-0.10Cr-0.12O) as Comparative Example 3 were subjected to ring pressing after oxidation quenching Experiments were carried out to observe its anti-embrittlement performance after high temperature oxidation quenching (anti-embrittlement performance under LOCA accident conditions).
  • the oxidation quenching process is as follows: the test temperature is 1204 °C, and the temperature is kept for a certain period of time.
  • the compensation strain value obtained by the hoop compression test after oxidative quenching reflects the residual plasticity of the test material after quenching (that is, the anti-embrittlement performance after high-temperature oxidative quenching). From the data shown in Table 3, it can be seen that the total content of iron and vanadium in Examples 3 and 4 Compared with the total content of iron and vanadium in Examples 1 and 2, the compensation strain is significantly lower, so the embrittlement resistance after high temperature oxidation quenching is not as good as that in Examples 1 and 2, but both are better than the Zr-4 alloy in Comparative Example 3. Compared with Zr-4 alloy, it has excellent resistance to embrittlement after high temperature oxidation quenching.
  • the zirconium alloys in the present invention within the content range of each component of the present invention and satisfying the content relationship have excellent corrosion resistance and embrittlement resistance after high temperature oxidation quenching, and are suitable for nuclear power plant reactors. cladding material.

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  • Engineering & Computer Science (AREA)
  • Metallurgy (AREA)
  • Organic Chemistry (AREA)
  • Physics & Mathematics (AREA)
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  • Crystallography & Structural Chemistry (AREA)
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Abstract

本发明公开了一种锆合金及其制作方法、包壳管及燃料组件，锆合金包括以下质量百分比的成分：铌0.45％～0.95％、锡0.21％～0.35％、铁0.03％～0.1％、钒0.03％～0.1％，还包括氧1000ppm-1600ppm，余量为锆；其中，铁和钒的总量≤0.15％。本发明的锆合金，通过各成分的配比，具有优良的耐腐蚀性能和高温氧化淬火后抗脆化性能好，适用于核电站反应堆的包壳材料，提高燃料组件的服役性能和安全性。

Description

锆合金及其制作方法、包壳管及燃料组件 技术领域
本发明涉及核燃料技术领域，尤其涉及一种锆合金及其制作方法、包壳管及燃料组件。
背景技术
核燃料组件用锆合金发展到现在，已经迭代了三代商用锆合金。目前使用的锆合金在具有较强抗蠕变性能的同时，在耐腐蚀性能方面存在不足。
为满足燃料不断提高的需求，不能只在某一性能方面进行优化设计，不仅要考虑常规工况，而且也需要考虑事故工况。随着锆合金的发展，核工业界对锆合金事故工况特别是失水事故下包壳行为的关注越来越高，特别是高温氧化淬火后的抗脆化能力。因此，有必要通过优化成分配比和添加其他元素可开发出在各种工况下性能都优良的锆合金。
发明内容
本发明要解决的技术问题在于，提供一种具有优良耐腐蚀性能和高温氧化淬火后抗脆化性能的锆合金及其制作方法、该锆合金制成的包壳管及具有该包壳管的燃料组件。
本发明解决其技术问题所采用的技术方案是：提供一种锆合金，包括以下质量百分比的成分：铌0.45％～0.95％、锡0.21％～0.35％、铁0.03％～0.1％、 钒0.03％～0.1％，还包括氧1000ppm-1600ppm，余量为锆；其中，铁和钒的总量≤0.15％。
优选地，所述锆合金中，C≤100ppm，N≤45ppm。
本发明还提供一种上述的锆合金的制作方法，包括以下步骤：
S1、提供分别含有锆、铌、锡、铁和钒成分的原料，根据锆合金中各成分所占的质量百分比称取原料；
S2、将所述原料进行熔炼，制得合金锭；
S3、将所述合金锭锻造成坯料；
S4、将所述坯料进行β淬火；
S5、将经过β淬火后的坯料进行多次冷轧，每次冷轧之间进行中间退火；
S6、将经过冷轧后的坯料进行最终退火，制得锆合金。
优选地，步骤S3中，所述锻造的温度为800℃-1100℃。
优选地，步骤S4中，所述β淬火的温度为950℃-1100℃。
优选地，步骤S5中，所述中间退火的温度为550℃-600℃。
优选地，步骤S5中，进行冷轧之前将所述坯料进行挤压或热轧。
优选地，步骤S6中，所述最终退火的温度为460℃-600℃。
优选地，所述制作方法还包括以下步骤：
S7、将所述锆合金加工成包壳管。
本发明还提供一种包壳管，采用以上任一项所述的锆合金制成。
本发明还提供一种燃料组件，包括以上所述的包壳管。
本发明的锆合金，通过各成分的配比，较于现有的Zr-4合金具有更优良的耐腐蚀性能和高温氧化淬火后抗脆化性能好，适用于核电站反应堆的包壳材料，提高燃料组件的服役性能和安全性。
具体实施方式
本发明的锆合金，为低锡锆锡铌铁钒合金，其包括以下质量百分含量的成分：
Nb(铌)0.45％～0.95％、Sn(锡)0.21％～0.35％、Fe(铁)0.03％～0.1％和V(钒)0.03％-0.1％，还包括O(氧)1000ppm～1600ppm，余量为Zr(锆)。其中，铁和钒的总量≤0.15％。
本发明的锆合金还包括：C(碳)≤100ppm、N(氮)≤45ppm。可以理解地，还包括一些不可避免且量少的杂质。
本发明的锆合金在1204℃氧化且氧化率达到18％时淬火后残余塑性值＞3.4％。
对于Nb(铌)，研究表明，锆合金中固溶铌对锆合金的耐腐蚀性能和抗蠕变性能都有好处，但铌的含量过高会对热处理敏感，因此本发明中，为保证锆合金具有优良的耐腐蚀性能和抗蠕变性能，Nb的含量控制在0.45wt％～0.95wt％，保证锆合金中铌的充分固溶。
锡(Sn)在锆中的固溶度较大，融入一定量的锡后，会提高锆合金的强度和抗蠕变性能，但是锡的添加会降低锆合金的耐均匀腐蚀能力。另一方面，锡的添加可以提高锆合金在高Li浓度环境下的耐腐蚀性能。本发明在确定锡含量时综合考虑了锡对耐腐蚀性能和对高Li环境下耐腐蚀性能的影响，将锡的含量控制在0.21wt％～0.35wt％，既提高锆合金对水化学的适应性，又最低程度的减少锡对耐腐蚀性能带来的不良影响，使锆合金具有优异的耐腐蚀性能。
铁(Fe)和钒(V)为过渡族金属元素，添加在锆合金中能够增加锆合金 的耐腐蚀性能，其中钒元素的添加可提高锆合金的抗吸氢性能。铁、钒的该类过渡族金属元素在锆合金中需要适量添加，添加过多时，会导致锆合金在高温氧化淬火后抗脆化性能明显下降。因此，本发明中，严格控制了Fe、V等元素的含量，Fe含量控制在0.03wt％～0.1wt％，V含量控制在0.03wt％～0.1wt％，并且铁和钒在锆合金中的总量≤于0.15wt％，保证该合金在1204℃氧化且氧化率达到18％(使用Cathcart-Pawel公式(锆合金氧化规律)计算)时淬火后残余塑性值大于＞3.4％，保证锆合金有足够的高温氧化淬火后抗脆化性能。
本发明的锆合金中，氧(O)的加入能够提高锆合金的强度和抗蠕变性能，但随着氧含量的升高，锆合金的可加工性会降低，特别是抗冲压性能。因此，氧的含量控制在1000ppm-1600ppm。
本发明的锆合金的制作方法，可包括以下步骤：
S1、提供分别含有锆、铌、锡、铁和钒成分的原料，根据锆合金中各成分所占的质量百分比称取原料(配料计算)。
例如，其中的锆原料使用核级海绵锆。铌、锡、铁和钒元素以纯金属或中间合金的形式添加。
S2、将原料进行熔炼，制得合金锭。
将所有原料放入真空熔炼炉中进行熔炼，调节O、C和N的含量，最后制得合金锭。
S3、将合金锭在800℃-1100℃的温度下锻造成坯料。
S4、将坯料进行β淬火。
其中，β淬火的温度为950℃-1100℃，并保温足够长时间使坯料整体到达淬火温度。S5、将经过β淬火后的坯料进行多次冷轧，每次冷轧之间进行中 间退火。
其中，根据所要形成的锆合金形态(如管材等)在冷轧之前，将坯料进行挤压或热轧，再将坯料进行至少4道次冷轧。中间退火的温度为550℃-600℃。
S6、将经过冷轧后的坯料在460℃-600℃下进行最终退火，制得锆合金。
锆合金可根据应用产品需要制成型材、板材或管材。
例如，本发明的制作方法还包括以下步骤：
S7、将步骤S6制得的锆合金加工成包壳管，以用于燃料组件。
在一应用实施方式中，将上述的锆合金制成燃料组件的包壳管。
对于燃料组件，包括上述的锆合金制成的包壳管，还包括密封在包壳管内的燃料芯块。由于包壳管由上述的锆合金制成，较于常规Zr-4合金制成的包壳管更优良的耐腐蚀性能、高温氧化淬火后抗脆化性能好，进而提高燃料组件的服役性能和安全性。
以下通过具体实施例对本发明作进一步说明。
根据本发明的制作方法制得实施例1-实施例4的锆合金，实施例1-实施例4的锆合金中各成分含量如表1所示。
表1
Figure PCTCN2021117834-appb-000001
Figure PCTCN2021117834-appb-000002
将实施例1-实施例4及比较例1-2制得的锆合金及Zr-4合金(Zr-1.30Sn-0.20Fe-0.10Cr-0.12O)作为比较例3进行腐蚀试验。腐蚀试验在高压釜上开展，腐蚀条件为360℃/18.6MPa/去离子水，试验时间为130天。结果如下表2所示。
表2
实施例 腐蚀量(mg/dm 2)
1 44.62
2 49.15
3 50.21
4 46.12
比较例1 45.33
比较例2 45.96
比较例3 63.30
从表2所示数据可知，实施例1-4及比较例1-2的锆合金较于常规的Zr-4合金具有较高的耐腐蚀性能。
将实施例1-实施例4及比较例1-2制得的锆合金及作为比较例3的Zr-4合金(Zr-1.30Sn-0.20Fe-0.10Cr-0.12O)进行氧化淬火后环压试验，以观察其高温氧化淬火后抗脆化性能(LOCA事故工况下抗脆化性能)。氧化淬火过程为：试验温度为1204℃，保温一定时间，使样品的CP-ECR(通过 Cathcart-Pawel公式计算出的等效锆反应量)达到18％时，使样品在200s内缓冷到800℃，再进行淬火，针对氧化淬火后样品开展环向压缩试验，根据载荷位移曲线计算出的补偿应变值大于3.4％时，说明样品具有足够的塑性。结果如下表3所示。
表3
实施例 补偿应变(％)
1 8.1
2 9.6
3 6.3
4 5.5
比较例1 3.1
比较例2 3.4
比较例3 4.0
氧化淬火后环向压缩试验获得的补偿应变值反映试验材料淬火后的残余塑性(即高温氧化淬火后抗脆化性能)，从表3所示数据可知，实施例3、4的铁钒总含量较于实施例1、2的铁钒总含量高，补偿应变明显更低，因此高温氧化淬火后抗脆化性能不如实施例1、2，但均优于比较例3的Zr-4合金，较于Zr-4合金具有优异的高温氧化淬火后抗脆化性能。比较例1、2由于铁钒总含量超过0.15wt％，因此高温氧化淬火后抗脆化性能相比于实施例1-4及比较例3差。实施例1-4的铁钒总含量不超过0.15wt％，补偿应变值都＞3.4％，说明实施例1-4制得的锆合金具备优异的高温氧化淬火后抗脆化性能。
可以理解地，本发明除了上述各实施例外，在本发明各成分含量范围内及满足含量关系式的锆合金，均具有优异的耐腐蚀性能和高温氧化淬火后抗 脆化性能，适用做核电站反应堆包壳材料。
以上所述仅为本发明的实施例，并非因此限制本发明的专利范围，凡是利用本发明说明书内容所作的等效结构或等效流程变换，或直接或间接运用在其他相关的技术领域，均同理包括在本发明的专利保护范围内。

Claims (11)

  1. 一种锆合金，其特征在于，包括以下质量百分比的成分：铌0.45％～0.95％、锡0.21％～0.35％、铁0.03％～0.1％、钒0.03％～0.1％，还包括氧1000ppm-1600ppm，余量为锆；其中，铁和钒的总量≤0.15％。
  2. 根据权利要求1所述的锆合金，其特征在于，所述锆合金中，C≤100ppm，N≤45ppm。
  3. 一种权利要求1或2所述的锆合金的制作方法，其特征在于，包括以下步骤：
    S1、提供分别含有锆、铌、锡、铁和钒成分的原料，根据锆合金中各成分所占的质量百分比称取原料；
    S2、将所述原料进行熔炼，制得合金锭；
    S3、将所述合金锭锻造成坯料；
    S4、将所述坯料进行β淬火；
    S5、将经过β淬火后的坯料进行多次冷轧，每次冷轧之间进行中间退火；
    S6、将经过冷轧后的坯料进行最终退火，制得锆合金。
  4. 根据权利要求3所述的锆合金的制作方法，其特征在于，步骤S3中，所述锻造的温度为800℃-1100℃。
  5. 根据权利要求3所述的锆合金的制作方法，其特征在于，步骤S4中，所述β淬火的温度为950℃-1100℃。
  6. 根据权利要求3所述的锆合金的制作方法，其特征在于，步骤S5中，所述中间退火的温度为550℃-600℃。
  7. 根据权利要求3所述的锆合金的制作方法，其特征在于，步骤S5中， 进行冷轧之前将所述坯料进行挤压或热轧。
  8. 根据权利要求3所述的锆合金的制作方法，其特征在于，步骤S6中，所述最终退火的温度为460℃-600℃。
  9. 根据权利要求3-8任一项所述的锆合金的制作方法，其特征在于，还包括以下步骤：
    S7、将所述锆合金加工成包壳管。
  10. 一种包壳管，其特征在于，采用权利要求1或2所述的锆合金制成。
  11. 一种燃料组件，其特征在于，包括权利要求10所述的包壳管。
PCT/CN2021/117834 2021-04-08 2021-09-10 锆合金及其制作方法、包壳管及燃料组件 WO2022213544A1 (zh)

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