Creep studies for zircaloy life prediction in water reactors

Zirconium alloys, commonly used as cladding tubes in water reactors, undergo complex biaxial creep deformation. The anisotropic nature of these metals makes it relatively complex to predict their dimensional changes in-reactor. These alloys exhibit transients in creep mechanisms as stress levels change. The underlying creep mechanisms and creep anisotropy depend on the alloy composition as well as the thermomechanical treatment. The anisotropic biaxial creep of cold-worked and recrystallized Zircaloy-4 in terms of Hill’s generalized stress formulation is described, and the temperature and stress dependencies of the steady-state creep rate are reviewed. Predictive models that incorporate anelastic strain are used for transient and transients in creep. For more information, contact K.L. Murty, North Carolina State University, Department of Nuclear Engineering, Campus Box 7909, Raleigh, North Carolina 27695-7909; (919) 515-3657; fax (919) 515-5115; e-mail murty@ncsu.edu.     

Effect of temperature on corrosion and semiconducting properties of oxide films formed on M5 zirconium alloy

Nuclear fuel cladding for pressurised water reactors is commonly manufactured with zirconium alloys. The M5 alloy is a relatively new cladding material for in-reactor used with enhanced performance compared to traditional zircaloys. In this work, the influence of temperature on the corrosion resistance and semiconducting properties of the passive film formed on the M5 alloy in a borate buffer solution has been evaluated. The electrochemical behaviour of the zirconium alloy was assessed by potentiodynamic polarisation tests, electrochemical impedance spectroscopy and Mott–Schottky plots. The results indicated that the corrosion resistance of the M5 alloy decreased with temperature due to the formation of a less stable and more defective passive film. The Mott–Schottky approach used in combination with polarisation tests and impedance measurements was effective to reveal the protective state of the passive film on the M5 alloy.     

Zirconium-Niobium Alloys for Core Elements of Pressurized Water Reactors

The main characteristics of niobium-bearing zirconium alloys used for fabricating fuel element claddings of pressurized water reactors are considered. It is shown that the high corrosion and radiation resistance of zirconium parts is provided by the chemical composition, structure, and phase composition of the alloys. The Zr – Nb alloys developed in Russia provide reliable operation of fuel elements and fuel rod arrays in active reactors and serve as a basis for new modifications.     

锆合金疖状腐蚀研究综述

疖状腐蚀是沸水堆中锆合金表面经常发生的1种局部腐蚀现象，它的产生直接影响包壳管的使用寿命和反应堆的安全性，为了全面认识疖状腐蚀的发生、发展及其控制因素，本文总结了国内外疖状腐蚀研究方面的一些主要成果，介绍了疖状蚀斑的形貌、形成机理以及及影响因素。在形成机制方面，目前主要有KUWAE的氢积聚模型和周邦新的形核长大模型。在疗状腐蚀的影响因素方面，认为主要有表面影响、热处理影响、合金成分影响、第二组影响、辐照影响等。最后指出了提高材料抗疗状腐蚀性能的工艺措施：提高Fe Cr含量、降低Sn含量、昼减少淬火后的退火次数和退火温度、降低锆合金制品的表面粗糙可以有效提高锆合金的抗疖状腐蚀能力，最根本的措施还是使用含铌新锆合金。     

锆合金包壳表面涂层的制备进展

锆因其极低的中子吸收截面、较高的熔点和优良的耐腐蚀性等特点,在核技术领域得到大量应用,主要作为核燃料的包壳材料。2011年日本福岛核事故后,事故容错燃料(ATF)的开发成为研究热点,尤其着重提高包壳材料的抗高温氧化性,而在锆合金表面制备涂层是提高该能力的重要途径之一。评述了锆合金包壳表面涂层的种类、性能、制备方法及各种方法的特点与发展。指出激光熔覆、等离子喷涂和冷喷涂都有沉积速率快、涂层厚的特点,但涂层过厚将降低核燃料的中子经济性。激光熔覆和等离子喷涂制得的涂层内应力大,存在较多气孔甚至微裂纹。冷喷涂涂层的应力和气孔得到改善,但喷涂法都存在粉尘及噪声污染等问题。重点分析了磁控溅射法(MS)和电弧离子镀(AIP)两种物理气相沉积技术在包壳涂层制备中的应用现状、存在的问题及未来发展方向。指出磁控溅射法因沉积速率可控、涂层的内应力小及涂层组分可调整等优势而应用最广。电弧离子镀因涂层致密、结合力强而最具发展潜力。这为进一步促进锆合金表面涂层的制备与研究提供了参考。     

核燃料包壳锆合金表面涂层研究进展

锆合金表面涂层是提高核燃料包壳事故容错能力的重要途径之一。本文综述了锆合金表面涂层的研究进展,包括涂层种类、制备工艺、微观组织以及抗水蒸气氧化性能、耐腐蚀性能等,介绍了锆合金表面涂层种类选择的依据,探讨了涂层的制备工艺、微观组织与性能之间的关系,分析了当前研究中存在的若干问题及未来涂层的发展方向,为进一步促进核燃料包壳锆合金表面涂层的研究提供了有价值的参考。     

压水堆燃料包壳锆合金中第二相腐蚀行为研究进展

锆合金是目前唯一大规模商用的压水堆燃料包壳材料,其耐水侧腐蚀性能是影响核反应堆安全性与经济性的重要因素。微量合金元素(Fe、Nb等)主要以第二相的形式弥散分布在锆合金基体中,但可对锆合金的腐蚀行为产生显著影响。本文比较了不同锆合金中第二相的差异,综述了锆合金中典型第二相的腐蚀行为及其影响因素。分别比较了二元及三元第二相中主要合金元素Fe和Nb的腐蚀过程,总结了不同水化学条件下第二相腐蚀产物的差异及其对锆合金基体腐蚀行为的影响,并指出当前针对第二相腐蚀行为研究中存在的不足。最后,对锆合金第二相腐蚀行为研究趋势进行了展望,先进微观表征手段可进一步完善含Fe、Nb元素第二相的腐蚀机理研究,将为提高我国新型锆合金包壳材料的耐腐蚀性能提供理论参考。     

Corrosion Cracking of Zirconium Cladding Tubes (A Review). I. Methods of Study and Mechanisms of Fracture

Methods of study and criteria of evaluation of stress corrosion cracking (SCC) of zirconium alloys are generalized as applied to cladding tubes of nuclear reactors. Mechanisms of SCC in zirconium cladding tubes in iodine-bearing media (iodine vapors, solution of iodine in methanol, etc.) are described. Metallographic and fracture features of damage in these media are analyzed. Data on cracking rates and critical stress intensity factors are presented.__________Translated from Metallovedenie i Termicheskaya Obrabotka Metallov, No. 2, pp. 31 – 39, February, 2005.     

锆合金包壳表面涂层研究进展

耐事故燃料是一种满足反应堆更多安全裕量设计要求的新型燃料元件。锆合金表面涂层研究是耐事故燃料包壳发展的一个主要方向,致力于解决高温条件下锆水严重反应的问题。该包壳具有经济性好,易于实现商业化等优点。重点阐述了锆合金包壳表面涂层制备技术和一些应用性能的研究进展,制备技术包括涂层方法、涂层厚度和涂层成分等,应用性能主要包括高温氧化和辐照性能。详细分析了锆合金表面涂层研究需要考虑的四个关键问题,即涂层材料选择、涂层工艺选择、涂层质量表征以及涂层锆包壳关键应用性能研究。涂层材料既要满足耐高温氧化性能,又要满足堆内正常运行的相关性能要求;涂层工艺应能制备出结合力好且致密的薄膜,并考虑锆包壳管涂层过程的可实现性;针对锆包壳特殊的应用环境,涂层质量表征重点关注涂层的附着力和膜致密度;涂层包壳关键应用性能主要考虑高温氧化、腐蚀、抗热冲击和腐蚀性能。综合已有研究结果,指出MAX相和金属Cr是两种有应用前景的锆包壳涂层材料,电弧离子镀技术作为锆包壳涂层工艺有一定的发展潜力。     

核燃料包壳锆合金表面铬涂层研究进展

锆合金凭借其较低的热中子吸收截面、优异的抗辐照性能以及良好的核燃料相容性等优点,被广泛应用于压水堆燃料包壳.福岛核事故后,表面铬涂层改性的锆合金成为耐事故包壳材料的重点研究方向之一,被认为是短期内最有可能投入商业应用的技术.综述了近年来核燃料包壳锆合金表面铬涂层的研究成果.介绍了铬涂层在事故条件下和正常工况条件下的性能优势,分析了其与锆合金基体在热性能上的匹配特性,重点对比了现有的铬涂层制备方法的优缺点,包括激光熔覆、喷涂、物理气相沉积等.其中激光熔覆和喷涂技术具有沉积速度较快、工艺条件相对简单的特点,但涂层厚度和粗糙度偏高,均匀性较差.物理气相沉积技术制得的涂层综合性能好,不足之处是涂层沉积速率较低,沉积过程需要高真空环境.兼顾高质量和低成本且适合商业化生产的包壳管表面铬涂层制备工艺仍有待于深入研究.归纳了铬涂层的高温氧化失效机制,提出在高温氧化过程中,涂层的分层、残余铬层的消耗以及锆元素沿铬晶界的扩散,是产生氧快速扩散通道并最终导致涂层失效的主要原因.最后指出了当前研究中存在的若干问题及其解决措施,为包壳锆合金表面铬涂层的进一步研究提供参考.     

钇离子注入锆的动电位极化曲线研究

通过对锆表面进行不同剂量的钇离子注入及对电位极化曲线的测量，分析了钇离子注入对锆电化学行为的影响。结果表明：钇离子注入能明显地提高锆在酸性、中性及碱性环境中的耐腐蚀性能。用X光电子能谱(XPS)分析了离子注入样品表层的成分与价态，进而探讨了钇离子注入表面改性的机理。     

Crack initiation and propagation behavior of zirconium cladding under an environment of iodine-induced stress corrosion

Tests of iodine-induced stress corrosion cracking (ISCC) were carried out to elucidate the initiation and propagation of cracks in the claddings of zirconium alloys. Zircaloy-4 cladding and Nb-contained zirconium cladding were pressurized with and without a pre-cracked state at 350°C in an iodine environment. The results show that pitting nucleation and growth play an important role in initiating ISCC. Pits preferentially grow and agglomerate around the grain boundary, where the number of pits increases with the iodine concentration and the hoop stress of the claddings. A model of grain-boundary pitting coalescence and a model of pitting-assisted slip cleavage, which were proposed to clearly elucidate the crack initiation and propagation process under ISCC, produce reasonable results. The Nb-contained zirconium cladding exhibits higher ISCC resistance than Zircaloy-4 from the standpoint of a higher threshold stress-intensity factor and a lower crack propagation rate.     

Corrosion Cracking of Zirconium Cladding Tubes. A Review. 2. Effect of External Factors, Structure, and Properties of the Alloys

Data on the effect of various external factors (applied loads, iodine concentration, temperature, irradiation), structure, and properties (strength, state of the surface, residual stresses, and hydrogen charging) of zirconium alloys on the mechanisms of and resistance to stress corrosion cracking (SCC) of zirconium cladding tubes primarily in iodine-bearing media are presented.     

预变形对Zr-Sn-Nb合金淬火时效晶粒及析出相的影响

采用SEM附带的背散射电子通道衬度（ECC）像、二次电子（SE）像及能谱（EDS）分析技术,研究了β相水淬后预变形处理对Zr-Sn-Nb合金在时效过程中再结晶和第二相析出的影响规律.结果表明,未引入预变形直接时效时所得组织中再结晶晶粒尺寸粗大且形状不规则,第二相粒子尺寸差异也较大,其中尺寸大的第二相粒子为含Cu的Zr₃Fe,主要沿原β晶界分布;预变形后再时效的组织中再结晶晶粒显著细化且尺寸均匀,第二相粒子尺寸差异减小,大尺寸的Zr₃Fe粒子主要沿α再结晶晶界分布.无论有无预变形或时效时间长短,晶粒内部析出相均为弥散分布的小尺寸Zr（Fe,Cr,Nb）₂粒子.引入预变形会减弱沉淀相沿晶界析出和急剧长大的倾向,使锆合金的微观组织和第二相分布特征改变.     

Fracture in ferritic reactor steel—dynamic strain aging and cyclic loading

Ferritic steels commonly used for pressure vessels and reactor supports in light water reactors exhibit dynamic strain aging resulting in decreased ductility and toughness. In addition, recent work indicated decreased toughness during reverse-cyclicloading that has implications on reliability of these structures under seismic loading conditions. This paper summarizes the authors’ recent work on these aspects, along with synergistic effects of interstitial impurity atoms and radiation-induced point defects, which result in interesting beneficial effects of radiation exposure at appropriate temperature and strain-rate conditions. While the cyclic loading effects on toughness are studied in A516 steel, the dynamic strain aging and radiation-defect interactions were investigated on pure iron as well as several ferritic steels. In addition, studies on fast vs. total (thermal+fast) neutron spectra revealed unexpected results due to the influence of radiation exposure on source hardening component of the yield stress; grain size of pure iron plays a significant role in these effects. The paper concludes with future research needed to address these concerns. For more information, contact K.L. Murty, North Carolina State University, Box 7909, Raleigh, North Carolina 27695-7909; (919) 515-3657; fax (919) 515-5115; e-mail murty@eos.ncsu.edu.     

焊接及随后热处理对Zr-1Nb合金腐蚀抗力的影响

为了推进从俄罗斯引进的田湾VVER-1000型核电站用Zr-1Nb合金燃料包壳和端塞棒材的国产化进程，研究了焊接及随后的真空热处理对Z4-1Nb合金抗360℃，18．6MPa水和500℃，10.3MPa蒸汽腐蚀性能的影响。实验结果表明焊接严重恶化Zr-1Nb合金在2种介质中的腐蚀抗力，主要是因为焊区中出现了粗大的马氏体组织；随后的真空退火对腐蚀抗力的影响取决于退火温度和腐蚀实验温度。但即使在最佳温度退火也不能使焊接态的腐蚀抗力恢复到非焊接态的水平。     

Zusammenhang zwischen Gefüge und Korrosionsverhalten bei Zirkonium

Interrelationship between structure and corrosion behaviour of zirconium Due to plant failures caused by the break-down of zirconium grade 702 subjected to sulphuric acid the structure and corrosion behaviour of welded and as delivered specimens were tested for various heat treatments. After annealing for 0.5 to 9 hours at temperatures from 650 to 1000 deg centigrade the specimens were cooled down in water, air, or furnace. The corrosion resistance was tested intentionally under extremely critical conditions, i.e. in boiling 65 pct sulphuric acid. It was shown by structure investigations and electron microprobe analysis that the corrosion behaviour of zirconium is strongly influenced by the structure, which in its turn is dependent on the grade of purity and the prehistory of the material. Type, amount, and distribution of residual elements or precipitations caused by them are responsible for the corrosion resistance. This is valid particularly for the element iron. The plant failures mentioned here coincided with the examination results. Measures to improve the chemical resistance of pure zirconium subjected to extremely aggressive media were derived as follows:

• 1 Annealing of the welded and as delivered state for several hours at 750 deg centigrade
• 2 If an improvement of the weld meets the demands welding should be conducted in at least two layers. The first layer should be in contact with the aggressive environment
• 3 Application of zirconium with a superior grade of purity as well for the base metal as for the filler material.

Practicability, economy, and success of the measures mentioned depend on the individual case to a high degree. In this regard practical experiences still have to be gained.     

Nanosized polycrystalline diamond cladding for surface protection of zirconium nuclear fuel tubes

A 300 nm thick polycrystalline diamond layer has been used for protection of zirconium alloy nuclear fuel cladding against undesirable oxidation with no loss of chemical stability and preservation of its functionality. Deposition of polycrystalline diamond layer was carried out using microwave plasma enhanced chemical vapor deposition apparatus with linear antenna delivery (which enables deposition of PCD layers over large areas). Polycrystalline diamond coated zirconium alloy fuel tubes were subjected to corrosion tests to replicate nuclear reactor conditions, namely irradiation and hot steam oxidation. Stable radiation tolerance of the polycrystalline diamond layer and its protective capabilities against hot steam oxidation of the zirconium alloy were confirmed. Finally, the use of polycrystalline diamond layers as a sensor of specific conditions (temperature/pressure dependent phase transition) in nuclear reactors is suggested.     

New insight into crack formation during corrosion of zirconium-based metal-oxide systems

Zirconium alloys are typically used in nuclear pressurized water reactors (PWR) as fuel cladding tubes due to their chemical stability and their mechanical strength at operating temperatures (≈300 °C). However, the corrosion of Zr-based cladding tubes is one of the factors limiting the burn-off rate in PWRs. It is commonly accepted that the corrosion kinetics involves a periodic succession of growth where the oxide thickness varies parabolically with time. As the oxide thickens, a cracking structure forms. The oxide appears striated with periodic layers of cracks running parallel to the metal/oxide interface. This cracking structure has been experimentally related to the periodicity of the oxide growth. In the present work, a finite-element study is used to investigate the development of stresses in the oxide under the combined influence of molar volume expansion during oxide formation, metal/oxide interface geometry and metallic substrate creep. The generation of tensile stresses capable of initiating the cracks that are observed experimentally is explored.     

改善锆合金疖状腐蚀的措施

锆合金以其独特的物理性能被广泛用于核反应堆堆芯结构材料,随着当前核电进一步向大功率、高燃耗发展,改善锆合金耐腐蚀性能便成为当务之急.介绍了改善锆合金疖状腐蚀的几种常用途径:优化合金成材过程中热加工制度;调节合金元素含量以及化学成分;对合金表面进行特殊工艺处理等.并在此基础上展望了锆合金抗疖状腐蚀技术的发展前景.