锆合金变形机理及其板材织构演化规律

锆是核工业的重要结构材料,又是优秀的化工耐蚀结构材料.锆合金的织构会对它的屈服强度、蠕变和疲劳强度、应力腐蚀开裂行为以及辐照尺寸变化等产生很大影响,因此变形机理的研究和织构控制在锆合金的开发利用中有十分重要的地位.综述了锆合金的变形机理,介绍了锆合金板材在不同轧制温度下的织构演化规律,以及退火温度对锆合金板材织构的影响,并总结了织构对锆合金板材力学性能的影响.最后指出,目前对锆合金板材加工后的织构进行精确预测还十分困难,需进行详细深入的研究,同时在加工中产生的织构对加工过程的影响以及与温度、应力分布、合金成分和组织的关系还需进一步认识.     

锆合金的织构及其对性能的影响

锆合金被广泛应用于核反应堆中，作为燃料元件的包壳材料或堆内结构部件，织构会影响其众多的应用性能，因此织构的研究及控制在锆合金的开发利用中具有重要作用。综述了，锆合金塑性变形的滑移和孪生系统；锆合金管板材的织构特点及控制方法；热轧温度对锆合金板材织构的影响以及退火处理锆合金织构的演化；并分别总结了织构与锆合金屈服强度、蠕变强度、碘致应力腐蚀开裂、氢化物取向分布以及辐照生长等性能的相互关系。     

成品退火温度对Zr-4锆合金管材变形织构和力学性能的影响

研究了退火温度对Zr-4锆合金成品金相组织、变形织构、应变收缩系数(CSR)及Kearns系数Fr的影响。结果表明:当成品管材在450~500℃区间进行消应力退火时,锆合金管材处于回复阶段,此过程不能减弱因金属形变导致晶面发生滑移形成的变形织构,与轧制态相比退火态合金管材CSR值无明显变化,Fr值随着退火温度的升高略增大;在500~530℃退火温度下,锆合金管材处于回复阶段与再结晶阶段的过渡区间,大部分变形织构转变为退火织构,Fr值和CSR值均急剧增大;在530~560℃退火温度下,锆合金管材处于再结晶阶段,组织为等轴状的再结晶形貌,原来的变形织构已基本完全被退火织构所替代,但此时锆合金管材还未达到完全再结晶,故Fr值和CSR值均略有升高。     

织构对先进锆合金蠕变性能各向异性的影响

研究了两种组织形貌相似的先进锆合金M5TM和N36锆合金核燃料包壳管材的单轴拉伸和内压蠕变性能.利用x射线衍射仪分析了它们的织构.试验发现两种先进锆合金包壳材料的蠕变性能表现出明显的各向异性.根据试验条件下的蠕变机理,结合弹性粘塑性自洽模型定性地分析了织构对锆合金管材蠕变各向异性的影响,解释了先进锫合金各向异性随应力指数变化的共性规律.揭示了织构与先进锆合金管材蠕变各向异性的定性关系.由于成分和织构因素的共同作用,在研究的试验条件下,N36合金的初始蠕变应变、稳态蠕变速率低于M5合金.织构是合金蠕变行为产生各向异性的主要原因,对于再结晶状态的先进锆合金包壳管,具有(0002)织构特征时,应力指数越高(即施加的应力水平越高),其蠕变的各向异性值越大.     

铝合金预拉伸板

性能优异、消除残余应力而机械加工或化铣后不变形的铝合金预拉伸板材,无论对航天、航空材料,还是各种用途的其他工业产品都占有极重要的地位.材料加工中,残余应力研究是一个带有方向性的领域,研究了板材在轧制过程和热处理淬火过程中,残余应力产生的机理和存在形式,并讨论了在拉伸机上,给予一定拉伸量的塑性变形后的板材,沿厚度在轧制方向上的残余应力重新分布,并向减少板材残余应力的趋势发展.     

连续退火对再加工锆合金腐蚀性能的影响

对在780℃下沿原板材横向热轧、冷轧及退火处理后的再加工锆合金进行金相织构、均匀腐蚀和疖状腐蚀测试和分析。结果表明:锆合金板材在再加工后及连续退火后,晶粒组织的粒度分布未发生明显变化,连续退火后的锆合金腐蚀性能未出现明显下降。     

连续退火对再加工锆合金腐蚀性能的影响

对在780℃下沿原板材横向热轧、冷轧及退火处理后的再加工锆合金进行金相织构、均匀腐蚀和疖状腐蚀测试和分析。结果表明:锆合金板材在再加工后及连续退火后,晶粒组织的粒度分布未发生明显变化,连续退火后的锆合金腐蚀性能未出现明显下降。     

Al-Mg合金预变形以及热处理对表面斜纹的影响

针对Al-Mg合金板材在变形过程中出现的表面斜纹,对比了Al-Mg合金与其他硬合金的应力-应变曲线特征差异,分析了表面纹路的形成机理以及形貌与应力-应变曲线的对应关系。根据不同应力-应变曲线形貌,开展了0%～5%不同预变形工艺、490～550℃范围加热、炉冷、水冷以及保温时间等退火工艺对材料性能和组织的影响研究。结果表明:退火后增加到2%以上的预变形可以有效解决屈服平台,但对锯齿状形貌没有改善。采用在510～540℃加热+水冷方式退火可以有效解决屈服平台和锯齿状形貌,并最终通过工业化验证,有效解决了表面斜纹问题。     

几种变形方式对钨合金组织性能及绝热剪切敏感性的影响

讨论了变形方式、变形量、变形材料的受力状态及微观组织结构等方面因素对材料性能和绝热剪切敏感性的影响。钨合金的受力状态、颗粒形状、微观组织取向对材料的变形、破坏和变形局域化机制有重要影响。X射线分析表明，旋锻后的钨合金组织有织构存在。力学性能的各向异性导致了材料绝热剪切破坏难易程度上的差异。     

5A30铝合金板高温拉伸本构关系研究

采用拉伸试样在Gleeble-1500材料热模拟试验机上对5A30铝合金进行高温拉伸实验,研究了该合金在变形温度为300～500℃,应变速率在0.01～10 s-1的高温流变变形行为。结果表明:变形温度和应变速率对该合金流变应力的大小有显著影响。流变应力随变形温度的升高而降低,随着应变速率的增加而升高。5A30铝合金的高温流变行为可用Zener-Hollomon参数描述,从流变应力、应变速率和变形温度的相关性,得出了该铝合金板材高温变形的材料常数和本构方程。计算出5A30铝合金板的变形激活能为Q=201.1 kJ.mol-1,材料常数为A=7.44×1013 s-1,n=4.3135,α=0.02 mm2.N-1;计算得到了5A30铝合金Arrhenius方程;利用双曲正弦模型,得到高温拉伸峰值应力和Z参数的解析式。     

Corrosion of Zirconium Alloys in Nuclear Applications - A Review

Abstract

With the recent demands for higher fuel discharge burn up towards better fuel utilisation and higher heat ratings and coolant temperatures for better thermal efficiency, interest on corrosion of zirconium alloys has been rekindled even though a large amount of work has been carried out to understand various aspects of corrosion of zirconium and its alloys in relation to their behaviour under nuclear operating conditions in the last 40 years or so. Though it seems that there exists some scope for further development of these alloys, which may need better understanding of the various corrosion mechanisms/ micromechanisms, in the present paper an attempt has been made to highlight a few corrosion aspects related to zirconium, however mainly zircaloys. Investigations carried out in this laboratory as well as elsewhere have been discussed keeping in view the aim of advancing the present knowledge so as to be able to meet the future requirements in terms of corrosion resistance as expected from this alloy system.     

Corrosion of Zirconium Alloys in Nuclear Applications – A Review

With the recent demands for higher fuel discharge burn up towards better fuel utilisation and higher heat ratings and coolant temperatures for better thermal efficiency, interest on corrosion of zirconium alloys has been rekindled even though a large amount of work has been carried out to understand various aspects of corrosion of zirconium and its alloys in relation to their behaviour under nuclear operating conditions in the last 40 years or so. Though it seems that there exists some scope for further development of these alloys, which may need better understanding of the various corrosion mechanisms/micromechanisms, in the present paper an attempt has been made to highlight a few corrosion aspects related to zirconium, however mainly zircaloys. Investigations carried out in this laboratory as well as elsewhere have been discussed keeping in view the aim of advancing the present knowledge so as to be able to meet the future requirements in terms of corrosion resistance as expected from this alloy system.     

Research progress of nuclear FeCrAl cladding materials

The leakage accident at the Fukushima nuclear power plant in Japan exposed the defect that zirconium alloy easily reacts with water vapor to cause an explosion, and distracting people's attention back to the FeCrAl alloy, which has relatively balanced properties. FeCrAl alloy has better corrosion resistance than zirconium alloy, and can better meet the working conditions of nuclear reactor under high temperature water vapor environment. In order to further ensure the safety of nuclear reactors, scholars from various countries had conducted a lot of research on the performance of FeCrAl alloy. For that reason, the corrosion mechanism of FeCrAl alloy and its corrosion behavior in high temperature steam and high temperature water environment were reviewed, and the effect of alloying elements (rare earth, Ti, Zr, Nb, and Mo), deformation and heat treatment system on the mechanical properties of the alloy was discussed; the research progress of the radiation and welding performance of FeCrAl alloy was summarized. The deficiencies in the research of nuclear grade FeCrAl alloys and the key development directions in the future were pointed out, which would provide references for other scholars in the field of research in the future.     

核级FeCrAl包壳材料研究进展

摘要：日本福岛核电站泄漏事故暴露了锆合金易与水蒸气反应造成爆炸的缺陷,又将人们的视线拉回了各项性能较为均衡的FeCrAl合金。FeCrAl合金较锆合金具有更好的耐腐蚀性能,更能满足核反应堆高温水蒸气环境下的工作条件。为进一步保障核反应堆安全性,各国学者对FeCrAl合金的性能进行了大量研究。基于此,综述了FeCrAl合金腐蚀机制及其在高温蒸汽和高温水环境中的腐蚀行为;探讨了合金元素（稀土、Ti、Zr、Nb和Mo）、变形和热处理对合金力学性能的影响;总结了FeCrAl合金辐照性能和焊接性能的研究进展。最后指出了核级FeCrAl合金在研究中存在的不足及未来重点发展方向,为今后其他学者在该领域的研究提供参考。     

改善Mg-Zn-Zr系变形合金铸锭塑性的研究

Mg-Zn-Zr系合全变形制品是当前使用较为广泛的高强轻合金材料,但是,因其合金铸锭塑性较差,故它的挤压制品、锻件及板材等得用一次挤压棒材和带板等,再通过二次压力加工方法来制得。本文研究了向Mg-Zn-Zr系合金中添加Y、Nd来改善合金铸锭塑性,从而达到用一次变形加工来获得挤压制品、锻件及板材方法的可行性问题,结果表明;在Mg-Zn-Zr系的MB15合金中添加1％Y和1％Nd,因其铸锭晶粒被细化,而可用铸锭直接制取挤压制品、锻件与板材,所获得的制品室温机械性能优于MB15合金,腐蚀性能则与MB15合金相当。     

SPD技术对锆及锆合金力学行为影响研究现状

综述了锆及锆合金剧烈塑性变形(SPD)后性能变化的研究进展,系统阐述了锆及锆合金经剧烈塑性变形后显微硬度、拉伸/压缩性能、高低周疲劳性能,重点介绍了SPD技术在纯锆、Zr-Nb系合金中的应用.经过剧烈塑性变形后,锆及锆合金的抗拉强度及屈服强度均显著提升,但依据剧烈塑性成形轨迹、合金成分、第二相分布、热处理制度不同,其提...     

近固溶限预析出对7055铝合金组织和力学性能的影响

通过改变固溶处理条件,即在近固溶限预析出并结合预析出前后的金相组织及不同时效制度下的电阻率和强度的对比,分析了近固溶限预析出对7055铝合金力学性能的影响。结果表明:在480℃时预析出在保持高的屈服强度和抗拉强度的同时改善和提高合金的抗SCC性能。     

Effect of corrosion damages on the mechanical properties of zirconium alloy cladding tubes

The accumulation kinetics of corrosion-induced damages in zirconium alloy cladding tubes is studied during stress corrosion cracking tests. The effect of corrosion damages on the mechanical properties of the tubes in loading by internal pressure is analyzed.