CN116121591A - Middle-strength high-plasticity Zr-4M zirconium alloy for cores - Google Patents
Middle-strength high-plasticity Zr-4M zirconium alloy for cores Download PDFInfo
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Abstract
本发明公开了一种核用中强高塑Zr‑4M锆合金，由以下质量百分比的元素成分组成：Sn 0.1%~1.4%，Fe 0.015%~0.22%，Cr 0.01%~0.11%，Hf 0.002%~0.01%，O 0.13%~0.4%，C 0.002%~0.01%，N 0.002%~0.006%，H 0.0005%~0.0015%，余量为Zr，其中Sn+O≤1.53%，Fe/Cr=1.5~3，O/H≥260，C/N=1~3。本发明通过调整Sn、Fe、Cr、Hf、C、N、H元素含量与比值结合对O成分设计，在不损失耐蚀性、中子吸收性能前提下，提升Zr‑4M锆合金的强度和塑性，满足了核反应堆元件盒用锆合金要求。The invention discloses a medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use, which is composed of the following element components in mass percentage: Sn 0.1%-1.4%, Fe 0.015%-0.22%, Cr 0.01%-0.11%, Hf 0.002 %~0.01%, O 0.13%~0.4%, C 0.002%~0.01%, N 0.002%~0.006%, H 0.0005%~0.0015%, the balance is Zr, of which Sn+O≤1.53%, Fe/Cr= 1.5~3, O/H≥260, C/N=1~3. In the present invention, by adjusting the content and ratio of Sn, Fe, Cr, Hf, C, N, H elements and combining the design of the O composition, the strength and the strength of the Zr-4M zirconium alloy are improved without loss of corrosion resistance and neutron absorption performance. Plasticity, which meets the requirements of zirconium alloys for nuclear reactor component boxes.
Description
技术领域technical field
本发明属于核用稀有金属材料技术领域，具体涉及一种核用中强高塑Zr-4M锆合金。The invention belongs to the technical field of rare metal materials for nuclear use, and in particular relates to a medium-strength and high-plasticity Zr-4M zirconium alloy for nuclear use.
背景技术Background technique
核反应堆元件作为核反应堆的心脏，元件中的锆合金凭借其低热中子吸收截面，是核反应堆元件用包壳、元件盒等结构材料的首选材料。锆合金是防止放射性物质泄漏的第一道屏障，对于维持核反应堆的连续正常运行、提高反应堆的服役可靠性与安全性具有关键作用。Nuclear reactor components are the heart of nuclear reactors. Zirconium alloys in components are the preferred materials for structural materials such as cladding and component boxes for nuclear reactor components due to their low thermal neutron absorption cross-section. Zirconium alloy is the first barrier to prevent the leakage of radioactive substances, and plays a key role in maintaining the continuous normal operation of nuclear reactors and improving the service reliability and safety of reactors.
核反应堆元件用锆合金的服役性能主要包括耐腐蚀性、力学性能、中子吸收性能等。当前，商用核反应堆元件用锆合金主要为Zr-4合金，其耐蚀性、力学性能、中子吸收性能均能满足常规使用要求，原因是核反应堆元件为固定装置，不会承受剧烈的力学加载或塑性变形。然而，对于核反应堆元件盒用锆合金，由于元件盒主要用于维持核反应堆元件的几何稳定性，要求元件盒用锆合金具有比常规锆合金更高的强度。此外，由于元件盒需要采用弯曲等加工方法进行加工，且元件盒在服役过程会发生振动或塑性变形，因此元件盒用锆合金必须同时具有较高的强度与塑性以免在服役或加工时发生脆性断裂。根据相关使用要求，元件盒用锆合金在保证不损失耐蚀性与中子吸收性能的前提下，不但需要将室温抗拉强度由现有商用锆合金的400MPa~450MPa提升到530MPa以上，而且需要将室温断后伸长率由现有商用锆合金的10%~15%左右提升到25%以上。The service performance of zirconium alloys for nuclear reactor components mainly includes corrosion resistance, mechanical properties, neutron absorption properties, etc. At present, zirconium alloys used in commercial nuclear reactor components are mainly Zr-4 alloys, and their corrosion resistance, mechanical properties, and neutron absorption properties can meet the requirements of conventional use. The reason is that nuclear reactor components are fixed devices and will not withstand severe mechanical loading or plastic deformation. However, for zirconium alloys for nuclear reactor component boxes, since the component boxes are mainly used to maintain the geometric stability of nuclear reactor components, zirconium alloys for component boxes are required to have higher strength than conventional zirconium alloys. In addition, since the component box needs to be processed by bending and other processing methods, and the component box will vibrate or plastically deform during service, the zirconium alloy used for the component box must have high strength and plasticity at the same time to avoid brittleness during service or processing fracture. According to relevant application requirements, on the premise of ensuring no loss of corrosion resistance and neutron absorption performance, zirconium alloys for component boxes not only need to increase the room temperature tensile strength from 400MPa~450MPa of existing commercial zirconium alloys to more than 530MPa, but also need to The elongation after fracture at room temperature is increased from about 10% to 15% of the existing commercial zirconium alloys to more than 25%.
核反应堆用锆合金必须具有较低的热中子吸收截面，以防止核裂变链式反应过程由于中子被吸收而终止。由于大多数元素的热中子吸收截面都远高于锆合金，并不适合于添加到核反应堆用锆合金当中，因此通过合金化对锆合金进行强塑性优化的难度极高。尽管添加中子吸收截面较低的合金元素，由于绝大多数合金元素在锆中的固溶度极低，所添加的合金元素会在锆合金中形成大量金属间化合物结构的第二相而严重损失锆合金的耐蚀性与断后伸长率。此外，现有商用锆合金的抗拉强度、断后伸长率距离目标值的距离均较大，利用传统方法提升强度的同时势必损失合金的断后伸长率，因此难以利用任何现有技术实现抗拉强度与断后延伸率的同步显著提升。Zirconium alloys for nuclear reactors must have a low thermal neutron absorption cross section to prevent the nuclear fission chain reaction process from being terminated due to neutron absorption. Since the thermal neutron absorption cross section of most elements is much higher than that of zirconium alloys, they are not suitable for adding to zirconium alloys for nuclear reactors, so it is extremely difficult to optimize the strong plasticity of zirconium alloys through alloying. Although the addition of alloying elements with low neutron absorption cross-sections, due to the extremely low solid solubility of most alloying elements in zirconium, the added alloying elements will form a large number of second phases of intermetallic compound structures in zirconium alloys and seriously The corrosion resistance and elongation after fracture of the zirconium alloy are lost. In addition, the tensile strength and elongation after fracture of existing commercial zirconium alloys are far away from the target value. When the strength is increased by traditional methods, the elongation after fracture of the alloy is bound to be lost. Therefore, it is difficult to use any existing technology to realize the tensile strength. The synchronization of tensile strength and elongation after fracture is significantly improved.
人们迫切希望获得一种核反应堆元件盒用中强高塑锆合金。People are eager to obtain a medium-strength and high-plastic zirconium alloy for nuclear reactor element boxes.
发明内容Contents of the invention
本发明所要解决的技术问题在于针对上述现有技术的不足，提供一种核用中强高塑Zr-4M锆合金。该方法通过调整现有Zr-4锆合金中Sn、Fe、Cr、Hf、C、N、H元素含量与相关比值，结合将间隙元素O作为合金元素进行成分设计，实现了在不损失耐蚀性、中子吸收性能的前提下，显著同步提升了Zr-4M锆合金的强度和塑性，解决了现有锆合金强度与延伸率无法同步提升的难题。The technical problem to be solved by the present invention is to provide a medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use in view of the above-mentioned deficiencies in the prior art. By adjusting the content and relative ratio of Sn, Fe, Cr, Hf, C, N, H elements in the existing Zr-4 zirconium alloy, combined with the composition design of the interstitial element O as an alloy element, the method realizes the corrosion resistance without loss Under the premise of high stability and neutron absorption performance, the strength and plasticity of Zr-4M zirconium alloy are significantly improved simultaneously, which solves the problem that the strength and elongation of existing zirconium alloys cannot be improved simultaneously.
为解决上述技术问题，本发明采用的技术方案为：一种核用中强高塑Zr-4M锆合金，其特征在于，由以下质量百分比的元素成分组成：Sn 0.1%~1.4%，Fe 0.015%~0.22%，Cr0.01%~0.11%，Hf 0.002%~0.01%，O 0.13%~0.4%，C 0.002%~0.01%，N 0.002%~0.006%，H0.0005%~0.0015%，余量为Zr，其中Sn与O的质量百分比总和不超过1.53%，且元素成分的质量百分比比值满足：Fe/Cr=1.5~3，O/H≥260，C/N=1~3。In order to solve the above-mentioned technical problems, the technical solution adopted in the present invention is: a medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use, characterized in that, it is composed of the following elemental components in mass percentages: Sn 0.1% ~ 1.4%, Fe 0.015 %~0.22%, Cr0.01%~0.11%, Hf 0.002%~0.01%, O 0.13%~0.4%, C 0.002%~0.01%, N 0.002%~0.006%, H0.0005%~0.0015%, remainder The amount is Zr, where the sum of the mass percentages of Sn and O does not exceed 1.53%, and the mass percentage ratios of the element components meet: Fe/Cr=1.5~3, O/H≥260, C/N=1~3.
本发明的研究过程中发现，在锆合金中Fe的扩散速率要高于Cr，从而在锆合金中倾向于形成ZrFe2或面心立方C15结构的Zr(Fe,Cr)2第二相，该结构与锆合金基体的HCP结构在晶格参数、晶体结构方面存在较大差异，界面处错配度较差，因此此类第二相会引起严重位错塞积，无法起到传递塑性变形的功能，对锆合金的塑性具有恶化作用。对此，本发明的Zr-4M锆合金通过控制Fe/Cr=1.5~3，有效抑制了面心立方C15结构的Zr(Fe,Cr)2第二相的形成，并有利于形成与锆合金基体相同晶体结构的密排六方C14结构Zr(Fe,Cr)2第二相的析出，同时由于Fe、Cr的相互置换作用，也避免了腐蚀电位较低的、恶化锆合金耐蚀性的ZrFe2第二相的析出，从而提高了Zr-4锆合金的塑性。In the research process of the present invention, it is found that the diffusion rate of Fe in zirconium alloys is higher than that of Cr, so that ZrFe2 or face-centered cubic C15 structure tends to be formed in zirconium alloys Zr (Fe, Cr) 2 second phases, the There are large differences in lattice parameters and crystal structure between the HCP structure and the HCP structure of the zirconium alloy matrix, and the mismatch degree at the interface is poor. Function, which has a deteriorating effect on the plasticity of zirconium alloys. In this regard, the Zr-4M zirconium alloy of the present invention effectively suppresses the formation of the second phase of Zr(Fe,Cr) of the face-centered cubic C15 structure by controlling Fe/Cr=1.5~3, and is conducive to the formation of the second phase of the Zr-4M zirconium alloy. The precipitation of the second phase of the hexagonal C14 structure Zr(Fe,Cr) 2 with the same crystal structure as the matrix, at the same time, due to the mutual replacement of Fe and Cr, ZrFe, which has a low corrosion potential and deteriorates the corrosion resistance of zirconium alloys, is also avoided. 2 The precipitation of the second phase improves the plasticity of the Zr-4 zirconium alloy.
其次，本发明的Zr-4M锆合金通过将O这一间隙元素作为合金化元素进行成分设计，并提升其含量，一方面，O自生具有较低的热中子吸收截面，不会恶化锆合金的中子吸收性能，另一方面，Zr-4M锆合金在有氧环境服役前其内部已经固溶部分O元素，能降低Zr-4M锆合金与有氧环境之间的O浓度梯度，减缓有氧环境中O的进入，进而降低其氧化增重速率，有利于锆合金在核反应堆内服役时在其表面自发形成致密钝化膜，提升了锆合金的耐蚀性；此外，O在Zr中的固溶度远高于Fe、Cr等合金元素，其添加不会形成金属间化合物结构的第二相，相反起到固溶强化作用，且不会损失较多塑性。因此，本发明的Zr-4锆合金通过提高O含量，使其耐蚀性和强度均得到提高，并保证其耐蚀性和塑性。Secondly, the Zr-4M zirconium alloy of the present invention is designed by using the interstitial element O as an alloying element to carry out composition design and increase its content. On the one hand, O self-generation has a lower thermal neutron absorption cross section and will not deteriorate the zirconium alloy. On the other hand, before the Zr-4M zirconium alloy served in the aerobic environment, some O elements have been solid-dissolved inside, which can reduce the O concentration gradient between the Zr-4M zirconium alloy and the aerobic environment, and slow down the The entry of O in the oxygen environment reduces its oxidation weight gain rate, which is conducive to the spontaneous formation of a dense passivation film on the surface of the zirconium alloy when it serves in a nuclear reactor, which improves the corrosion resistance of the zirconium alloy; in addition, the O in Zr The solid solubility is much higher than that of alloying elements such as Fe and Cr, and its addition will not form the second phase of the intermetallic compound structure, but will act as a solid solution strengthening effect without losing more plasticity. Therefore, the Zr-4 zirconium alloy of the present invention improves its corrosion resistance and strength by increasing the O content, and ensures its corrosion resistance and plasticity.
Sn与O元素在Zr基体中的溶解度较高，是Zr-4M锆合金的主要固溶强化元素，其中Sn的原子半径与Zr接近，为置换固溶强化，O的原子半径显著小于Zr，为间隙固溶强化，这两种固溶强化机制都是基于对Zr基体晶格上的位点进行不同程度的置换和间隙填充，因此具有较强竞争性。当两者强化程度都较高时，会导致Zr的晶格发生严重畸变，不利于位错滑移、孪生等塑性变形，进而导致锆合金的断后伸长率显著降低。因此，本发明的Zr-4M锆合金中通过限定Sn与O的质量百分比总和不超过1.53%，以避免过度固溶强化导致的Zr-4M锆合金塑性下降。本发明的研究过程中发现当Sn与O的质量百分比总和不超过1.53%时，在保证Zr-4M合金达到所需强度的前提下，同时保证了较高的断后伸长率。此外，Sn元素自身含量过高，会导致Sn元素的局部偏析，引起疖状腐蚀等非均匀腐蚀，降低了Zr-4M锆合金的耐蚀性，故本发明限定Sn 的质量百分比含量上限为1.4%。The solubility of Sn and O elements in the Zr matrix is high, and they are the main solid solution strengthening elements of the Zr-4M zirconium alloy. The atomic radius of Sn is close to that of Zr, which is a replacement solid solution strengthening. The atomic radius of O is significantly smaller than that of Zr, which is Interstitial solid solution strengthening, these two solid solution strengthening mechanisms are based on different degrees of substitution and interstitial filling of sites on the Zr matrix lattice, so they are highly competitive. When the strengthening degree of both is high, it will lead to serious distortion of the Zr lattice, which is not conducive to plastic deformation such as dislocation slip and twinning, and then leads to a significant decrease in the elongation after fracture of the zirconium alloy. Therefore, in the Zr-4M zirconium alloy of the present invention, the sum of the mass percentages of Sn and O is limited to not exceed 1.53%, so as to avoid the decrease in plasticity of the Zr-4M zirconium alloy caused by excessive solid solution strengthening. During the research process of the present invention, it is found that when the sum of the mass percentages of Sn and O is not more than 1.53%, the Zr-4M alloy can ensure a higher elongation after fracture under the premise of ensuring that the Zr-4M alloy reaches the required strength. In addition, the high content of Sn element itself will lead to local segregation of Sn element, causing non-uniform corrosion such as boil corrosion, and reducing the corrosion resistance of Zr-4M zirconium alloy. Therefore, the present invention limits the upper limit of the mass percentage content of Sn to 1.4 %.
再次，本发明的Zr-4M锆合金通过降低Hf、Sn的含量以及恶化断后伸长率的C、N、H含量，并严格控制具有形成间隙固溶体相互竞争关系的O与H、C与N的比例，即O/H≥260，C/N=1~3，一方面进一步降低了锆合金的热中子吸收截面，使其更好地维持核裂变链式反应，满足核反应堆用要求，另一方面，有效提升了锆合金中可用于间隙固溶、置换固溶的原子间隙或置换位点，保证了Zr-4锆合金在进行合金化的同时具有极高的断后伸长率，改善了Zr-4锆合金的塑性。Again, the Zr-4M zirconium alloy of the present invention is by reducing the content of Hf, Sn and the content of C, N, H which deteriorates the elongation after fracture, and strictly controls the relationship between O and H, C and N which form interstitial solid solutions. The ratio, that is, O/H≥260, C/N=1~3, on the one hand further reduces the thermal neutron absorption cross section of the zirconium alloy, so that it can better maintain the nuclear fission chain reaction and meet the requirements for nuclear reactors, on the other hand On the one hand, it effectively improves the atomic interstitial or replacement sites that can be used for interstitial solid solution and replacement solid solution in zirconium alloys, ensuring that Zr-4 zirconium alloys have extremely high elongation after fracture while alloying, and improve the Zr-4 zirconium alloy. -4 Plasticity of zirconium alloys.
上述的一种核用中强高塑Zr-4M锆合金，其特征在于，由以下质量百分比的元素成分组成：Sn 1.0%~1.4%，Fe 0.1%~0.22%，Cr 0.05%~0.11%，Hf 0.002%~0.01%，O 0.13%~0.4%，C 0.002%~0.01%，N 0.002%~0.006%，H 0.0005%~0.0015%，余量为Zr，其中Sn与O的质量百分比总和不超过1.53%，且元素成分的质量百分比比值满足：Fe/Cr=1.5~3，O/H≥260，C/N=1~3。The above-mentioned medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use is characterized in that it is composed of the following elemental components in mass percentage: Sn 1.0%~1.4%, Fe 0.1%~0.22%, Cr 0.05%~0.11%, Hf 0.002%~0.01%, O 0.13%~0.4%, C 0.002%~0.01%, N 0.002%~0.006%, H 0.0005%~0.0015%, the balance is Zr, wherein the sum of the mass percentages of Sn and O does not exceed 1.53%, and the mass percentage ratio of the element composition satisfies: Fe/Cr=1.5~3, O/H≥260, C/N=1~3.
上述的一种核用中强高塑Zr-4M锆合金，其特征在于，所述Zr-4M锆合金在室温下的拉伸性能为：抗拉强度Rm≥530MPa，屈服强度Rp0.2≥460MPa，断后伸长率A≥25%。The above-mentioned medium-strength and high-plasticity Zr-4M zirconium alloy for nuclear use is characterized in that the tensile properties of the Zr-4M zirconium alloy at room temperature are: tensile strength Rm≥530MPa, yield strength Rp 0.2≥460MPa , Elongation after breaking A≥25%.
上述的一种核用中强高塑Zr-4M锆合金，其特征在于，所述Zr-4M锆合金在288℃下的拉伸性能为：抗拉强度Rm≥230MPa，屈服强度Rp0.2≥160MPa，断后伸长率A≥35%。The aforementioned medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use is characterized in that the tensile properties of the Zr-4M zirconium alloy at 288°C are: tensile strength Rm≥230MPa, yield strength Rp 0.2≥160MPa , Elongation after breaking A≥35%.
上述的一种核用中强高塑Zr-4M锆合金，其特征在于，所述Zr-4M锆合金的理论中子吸收截面小于0.242×10-24cm2。通常采用蒙特卡洛粒子输运方法对合金的热中子吸收截面进行理论计算。The aforementioned medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use is characterized in that the theoretical neutron absorption cross section of the Zr-4M zirconium alloy is less than 0.242×10 -24 cm 2 . The theoretical calculation of the thermal neutron absorption cross section of the alloy is usually carried out by Monte Carlo particle transport method.
本发明核用中强高塑Zr-4M锆合金的制备方法为：根据目标产物Zr-4M锆合金的成分组成准备原料，然后经真空自耗电弧熔炼或真空感应熔炼得到Zr-4M锆合金铸锭，再分别通过β相区锻造、热轧、冷轧、热处理工艺，获得所需外形及尺寸的Zr-4M锆合金型材。The preparation method of the medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use of the present invention is as follows: prepare raw materials according to the composition of the target product Zr-4M zirconium alloy, and then obtain the Zr-4M zirconium alloy through vacuum consumable arc melting or vacuum induction melting Ingot casting, and then through β-phase zone forging, hot rolling, cold rolling, and heat treatment processes to obtain Zr-4M zirconium alloy profiles of required shape and size.
本发明与现有技术相比具有以下优点：Compared with the prior art, the present invention has the following advantages:
1、本发明对现有Zr-4锆合金的化学成分进行重新设计，并调整不影响锆合金耐蚀性和中子吸收性能的元素含量，通过控制Fe、Cr元素含量及其相关比值，使得恶化合金强度与塑性的Zr(Fe,Cr)2第二相的含量下降，同时调整Sn、Fe、Cr、Hf、C、N、H元素含量以及O与H、C与N的比例，降低锆合金的热中子吸收截面并提高断后伸长率，结合提高O元素含量进行固溶强化，实现了在不损失耐蚀性、中子吸收性能的前提下，提高Zr-4M锆合金的强度和耐蚀性并显著提升其塑性，满足了核反应堆元件盒用锆合金的使用需求。1. The present invention redesigns the chemical composition of the existing Zr-4 zirconium alloy, and adjusts the content of elements that do not affect the corrosion resistance and neutron absorption performance of the zirconium alloy. By controlling the content of Fe and Cr elements and their relative ratios, the The content of the second phase of Zr(Fe,Cr) 2 that deteriorates the strength and plasticity of the alloy decreases, and at the same time adjust the content of Sn, Fe, Cr, Hf, C, N, H elements and the ratio of O to H, C to N, and reduce the content of zirconium The thermal neutron absorption cross section of the alloy and the elongation after fracture are improved, combined with the increase of the O element content for solid solution strengthening, the strength and Corrosion resistance and significantly improved plasticity meet the requirements for the use of zirconium alloys for nuclear reactor component boxes.
2、本发明基于Sn置换固溶+O间隙固溶协调控制、间隙固溶原子协同强化、间隙原子优先填充锆的四面体间隙、Fe/Cr电子浓度设计理论，在不调整现有Zr-4锆合金元素种类的前提下，通过对元素含量及相关元素比例进行控制，使得Zr-4M锆合金的强度和塑性得到显著同步提升，弥补了现有锆合金强度与延伸率无法同步提升的缺陷。2. The present invention is based on the coordinated control of Sn replacement solid solution + O interstitial solid solution, synergistic strengthening of interstitial solid solution atoms, interstitial atoms preferentially filling the tetrahedral gap of zirconium, and Fe/Cr electron concentration design theory, without adjusting the existing Zr-4 On the premise of the types of zirconium alloy elements, by controlling the content of elements and the proportion of related elements, the strength and plasticity of Zr-4M zirconium alloys are significantly improved simultaneously, which makes up for the defect that the strength and elongation of existing zirconium alloys cannot be improved synchronously.
3、本发明的Zr-4M锆合金通过控制Fe、Cr含量及比例，有效限定了密排六方C14结构Zr(Fe, Cr)2第二相的体积分数，使其维持在合理的范围内，在提供钉扎强化的基础上，不损失Zr-4M锆合金的塑性。3. The Zr-4M zirconium alloy of the present invention effectively limits the volume fraction of the second phase of the hexagonal C14 structure Zr(Fe, Cr) 2 by controlling the content and ratio of Fe and Cr, so that it is maintained within a reasonable range, On the basis of providing pinning strengthening, the plasticity of the Zr-4M zirconium alloy is not lost.
4、本发明的Zr-4M锆合金在室温下的抗拉强度Rm可达530MPa以上，屈服强度Rp0.2可达460MPa以上，断后伸长率A可达25%以上；在288℃服役温度下，Zr-4M合金抗拉强度Rm可达230MPa以上，屈服强度Rp0.2可达160MPa以上，断后伸长率A可达35%以上，且理论计算出的Zr-4M合金热中子吸收截面小于0.242×10-24cm2。4. The tensile strength Rm of the Zr-4M zirconium alloy of the present invention can reach more than 530MPa at room temperature, the yield strength Rp 0.2 can reach more than 460MPa, and the elongation A after fracture can reach more than 25%; at a service temperature of 288°C, The tensile strength Rm of Zr-4M alloy can reach more than 230MPa, the yield strength Rp 0.2 can reach more than 160MPa, the elongation A after fracture can reach more than 35%, and the theoretically calculated thermal neutron absorption cross section of Zr-4M alloy is less than 0.242× 10-24 cm 2 .
5、本发明的Zr-4M锆合金是在现有商用Zr-4锆合金的基础上进行的成分改良而得到，在完全覆盖商用Zr-4锆合金的应用场景的前提下，还能满足商用Zr-4锆合金无法满足的例如核反应堆元件盒等更为严苛的应用场景，应用范围广，实用价值大。5. The Zr-4M zirconium alloy of the present invention is obtained by improving the composition on the basis of the existing commercial Zr-4 zirconium alloy. Zr-4 zirconium alloys can not meet more severe application scenarios such as nuclear reactor component boxes, and have a wide range of applications and great practical value.
下面通过实施例对本发明的技术方案作进一步的详细描述。The technical solutions of the present invention will be described in further detail below through examples.
具体实施方式Detailed ways
实施例1Example 1
本实施例的核用中强高塑Zr-4M锆合金由以下质量百分比的元素成分组成：Sn0.1%，Fe0.015%，Cr 0.01%，Hf 0.002%，O 0.13%，C 0.002%，N 0.002%，H 0.0005%，余量为Zr，其中Sn与O的质量百分比总和为0.23%，且元素成分的质量百分比比值满足：Fe/Cr=1.5，O/H=260，C/N=1。The medium-strength and high-plasticity Zr-4M zirconium alloy of the present embodiment is made up of the following element composition of mass percent: Sn0.1%, Fe0.015%, Cr 0.01%, Hf 0.002%, O 0.13%, C 0.002%, N 0.002%, H 0.0005%, the balance is Zr, the sum of the mass percentages of Sn and O is 0.23%, and the mass percentage ratio of the element composition satisfies: Fe/Cr=1.5, O/H=260, C/N= 1.
本实施例的核用中强高塑Zr-4M锆合金的制备方法为：根据目标产物Zr-4M锆合金的成分组成准备原料，然后经真空自耗电弧熔炼或真空感应熔炼得到Zr-4M锆合金铸锭，再分别通过1100℃锻造、热轧、冷轧、热处理工艺以及表面修磨，获得厚度为2.0mm的Zr-4M锆合金板材。The preparation method of the medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use in this embodiment is as follows: prepare raw materials according to the composition of the target product Zr-4M zirconium alloy, and then obtain Zr-4M by vacuum consumable arc melting or vacuum induction melting Zirconium alloy ingots were cast, and then through forging at 1100 ° C, hot rolling, cold rolling, heat treatment and surface grinding to obtain Zr-4M zirconium alloy plates with a thickness of 2.0 mm.
实施例2Example 2
本实施例的核用中强高塑Zr-4M锆合金由以下质量百分比的元素成分组成：Sn1.4%，Fe0.22%，Cr 0.11%，Hf 0.01%，O 0.13%，C 0.006%，N 0.002%，H 0.0005%，余量为Zr，其中Sn与O的质量百分比总和为1.53%，且元素成分的质量百分比比值满足：Fe/Cr=2，O/H=260，C/N=3。The medium-strength and high-plasticity Zr-4M zirconium alloy of the present embodiment is made up of the following element composition of mass percentage: Sn1.4%, Fe0.22%, Cr 0.11%, Hf 0.01%, O 0.13%, C 0.006%, N 0.002%, H 0.0005%, the balance is Zr, wherein the sum of the mass percentages of Sn and O is 1.53%, and the mass percentage ratio of the element composition satisfies: Fe/Cr=2, O/H=260, C/N= 3.
本实施例的核用中强高塑Zr-4M锆合金的制备方法为：根据目标产物Zr-4M锆合金的成分组成准备原料，然后经真空自耗电弧熔炼或真空感应熔炼得到Zr-4M锆合金铸锭，再分别通过1100℃锻造、热轧、冷轧、热处理工艺以及表面修磨，获得厚度为2.2mm的Zr-4M锆合金板材。The preparation method of the medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use in this embodiment is as follows: prepare raw materials according to the composition of the target product Zr-4M zirconium alloy, and then obtain Zr-4M by vacuum consumable arc melting or vacuum induction melting Zirconium alloy ingots were cast, and then through forging at 1100 ° C, hot rolling, cold rolling, heat treatment and surface grinding to obtain Zr-4M zirconium alloy plates with a thickness of 2.2 mm.
实施例3Example 3
本实施例的核用中强高塑Zr-4M锆合金由以下质量百分比的元素成分组成：Sn1.0%，Fe0.06%，Cr 0.02%，Hf 0.005%，O 0.4%，C 0.01%，N 0.006%，H 0.0015%，余量为Zr，其中Sn与O的质量百分比总和为1.4%，且元素成分的质量百分比比值满足：Fe/Cr=3，O/H=266.66，C/N=1.66。The nuclear medium-strength and high-plasticity Zr-4M zirconium alloy of the present embodiment is made up of the element composition of following mass percentage: Sn1.0%, Fe0.06%, Cr 0.02%, Hf 0.005%, O 0.4%, C 0.01%, N 0.006%, H 0.0015%, the balance is Zr, the sum of the mass percentages of Sn and O is 1.4%, and the mass percentage ratio of the element composition satisfies: Fe/Cr=3, O/H=266.66, C/N= 1.66.
本实施例的核用中强高塑Zr-4M锆合金的制备方法为：根据目标产物Zr-4M锆合金的成分组成准备原料，然后经真空自耗电弧熔炼或真空感应熔炼得到Zr-4M锆合金铸锭，再分别通过1100℃锻造、热轧、冷轧、热处理工艺以及表面修磨，获得厚度为2.5mm的Zr-4M锆合金板材。The preparation method of the medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use in this embodiment is as follows: prepare raw materials according to the composition of the target product Zr-4M zirconium alloy, and then obtain Zr-4M by vacuum consumable arc melting or vacuum induction melting Zirconium alloy ingots were cast, and then through forging at 1100 ° C, hot rolling, cold rolling, heat treatment and surface grinding to obtain Zr-4M zirconium alloy plates with a thickness of 2.5 mm.
实施例4Example 4
本实施例的核用中强高塑Zr-4M锆合金由以下质量百分比的元素成分组成：Sn1.1%，Fe0.1%，Cr 0.05%，Hf 0.008%，O 0.37%，C 0.01%，N 0.006%，H 0.0008%，余量为Zr，其中Sn与O的质量百分比总和为1.47%，且元素成分的质量百分比比值满足：Fe/Cr=2，O/H=462.5，C/N=1.66。The medium-strength and high-plasticity Zr-4M zirconium alloy of the present embodiment is made up of the following element composition of mass percent: Sn1.1%, Fe0.1%, Cr 0.05%, Hf 0.008%, O 0.37%, C 0.01%, N 0.006%, H 0.0008%, the balance is Zr, the sum of the mass percentages of Sn and O is 1.47%, and the mass percentage ratio of the element composition satisfies: Fe/Cr=2, O/H=462.5, C/N= 1.66.
本实施例的核用中强高塑Zr-4M锆合金的制备方法为：根据目标产物Zr-4M锆合金的成分组成准备原料，然后经真空自耗电弧熔炼或真空感应熔炼得到Zr-4M锆合金铸锭，再分别通过1100℃锻造、热轧、冷轧、热处理工艺以及表面修磨，获得厚度为2.5mm的Zr-4M锆合金板材。The preparation method of the medium-strength and high-plastic Zr-4M zirconium alloy for nuclear use in this embodiment is as follows: prepare raw materials according to the composition of the target product Zr-4M zirconium alloy, and then obtain Zr-4M by vacuum consumable arc melting or vacuum induction melting Zirconium alloy ingots were cast, and then through forging at 1100 ° C, hot rolling, cold rolling, heat treatment and surface grinding to obtain Zr-4M zirconium alloy plates with a thickness of 2.5 mm.
对比例1Comparative example 1
本对比例的商用Zr-4锆合金由以下质量百分比的元素成分组成：Sn 1.28%，Fe0.11%，Cr 0.11%，Hf 0.01%，O 0.12%，C 0.027%，N 0.008%，H 0.0025%，余量为Zr。The commercial Zr-4 zirconium alloy of this comparative example is composed of the following elements by mass percentage: Sn 1.28%, Fe0.11%, Cr 0.11%, Hf 0.01%, O 0.12%, C 0.027%, N 0.008%, H 0.0025 %, the balance is Zr.
经检测，本发明实施例1~4的Zr-4M锆合金与对比例1的商用Zr-4锆合金的室温及高温288℃拉伸性能数据如下表1所示。After testing, the room temperature and high temperature 288°C tensile performance data of the Zr-4M zirconium alloys of Examples 1-4 of the present invention and the commercial Zr-4 zirconium alloy of Comparative Example 1 are shown in Table 1 below.
从表1可知，相较于商用Zr-4锆合金，本发明的Zr-4M锆合金的室温及高温288℃拉伸性能包括抗拉强度、屈服强度、断后伸长率同时得到了显著提升。It can be seen from Table 1 that compared with the commercial Zr-4 zirconium alloy, the room temperature and high temperature 288°C tensile properties of the Zr-4M zirconium alloy of the present invention, including tensile strength, yield strength, and elongation after fracture, have been significantly improved at the same time.
以上所述，仅是本发明的较佳实施例，并非对本发明作任何限制。凡是根据发明技术实质对以上实施例所作的任何简单修改、变更以及等效变化，均仍属于本发明技术方案的保护范围内。The above descriptions are only preferred embodiments of the present invention, and do not limit the present invention in any way. All simple modifications, changes and equivalent changes made to the above embodiments according to the technical essence of the invention still belong to the protection scope of the technical solution of the invention.
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