锆合金疖状腐蚀研究进展

简要综述了近年来国内外关于锆合金疖状腐蚀的研究进展,归纳总结了锆合金的疖状腐蚀机理,主要包括氢聚集腐蚀、局部溶质贫化腐蚀、形核长大腐蚀、疖状斑形成腐蚀等。分析了影响锆合金耐疖状腐蚀的因素,并在此基础上提出了改善锆合金耐疖状腐蚀性能的方法,主要有：调节合金元素种类及含量;表面晶粒细化处理;进行适当的热处理,使第二相尺寸细化,分布均匀。     

西北锆管公司成功解决锆管疖状腐蚀

Zr-4合金包壳管的抗疖状腐蚀问题是锆锡合金的最大技术问题之一。西北锆管有限责任公司在各攻关单位及参研人员密切协作大力配合下，进行了2个阶段的试验研究。第一阶段主要是工模具设计制造、设备改造、工艺探索及确定疖状腐蚀试验规范；第二阶段主要是工业规模试验暨首件包壳管的生产与性能检验，生产的成品管其尺寸精度，表面状态，无损检验，疖状腐蚀及其它理化性能经设计、科研、生产、质保、质检等方面专家的联合现场检查，全部性能指标都满足了核用锆材的技术条件要求，尤其是抗疖状腐蚀性能得到了显著提高，成功地解决了Zr-4合金包…     

激光表面处理对Zr－4合金板材疖状腐蚀的影响

研究了几种激光表面处理工艺对不同状态的Ｚｒ－４合金板材泡疖腐蚀的影响。实验结果表明：激光表面处理可以显著提高Ｚｒ－４合金抗疖状腐蚀性能；激光表面处理前Ｚｒ－４合金板材的状态（消除应力退火态、再结晶退火态、冷加工态）对激光表面处理后板材的疖状腐蚀性能的影响不明显     

激光表面处理对Zr—4合金板材疖状腐蚀的影响

研究了几种激光表面处理工艺对不同状态的Ｚｒ－４合金板材泡疖腐蚀的影响。实验结果表明：激光表面处理可以显著提高Ｚｒ－４合金抗疖状腐蚀性能；激光表面处理前Ｚｒ－４合金板材的状态（消除应力退火态、再结晶退火态、冷加工态）对激光表面处理后板材的疖状腐蚀性能的影响不明显     

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.     

锡含量对锆－锡合金性能的影响

研究了锡含量对锆－锡合金力学性能和腐蚀性能的影响。结果表明：锡在０．５ｗｔ％～１．７ｗｔ％范围内，随着锡含量的增加，合金的强度增加，在４００℃蒸汽腐蚀试验中，随着合金中锡含量的减少，腐蚀增重降低并能改善合金的抗疖状腐蚀能力。     

锆—4合金的疖状腐蚀

对不同状态的锆-4合金管材进行了高温蒸汽腐蚀研究。经400℃预膜和未预膜的样品在460℃/10.3 MPa的蒸汽中进行腐蚀试验,结果表明:未预膜的样品迅速腐蚀,而预膜处理延缓了疖状腐蚀的产生;不同加工工艺的样品的腐蚀结果截然不同。挤压前进行β淬火使Zr-4合金的抗疖状腐蚀性能得到改善,应变时效处理后的Zr-4合金管材具有良好的抗疖状腐蚀性能。疖状腐蚀的敏感性与合金的组织状态密切相关,取决于合金中的第二相粒子的数密度、平均粒子大小和基体中溶质元素的含量。     

热处理影响Zr-4合金耐疖状腐蚀性能的机制

采用500℃，10．3MPa过热蒸汽腐蚀方法，研究了热处理对Zr-4合金耐疖状腐蚀性能的影响。试样经过600，820和1000℃不同热处理后，耐疖状腐蚀性能明显不同。提高Fe，Cr合金元素在α-Zr中过饱和固溶含量，可以明显改善耐疖状腐蚀性能，第二相的大小不是决定的因素。用高分辨扫描电镜观察了氧化膜的内表面形貌和断口形貌，研究了耐疖状腐蚀性能与氧化膜显微组织之间的关系。从疖状腐蚀斑的成核与长大，热处理会引起Fe和Cr合金元素在α-Zr中过饱和固溶含量的变化，以及从氧化膜生长的各向异性与α-Zr中合金元素过饱和固溶含量的关系出发，讨论了热处理影响耐疖状腐蚀性能的机制。     

Zr-4合金氧化膜显微组织与疖状腐蚀机制研究

经过不同热处理后的几种Zr-4合金样品,在550 ℃/25 MPa超临界水中腐蚀时都不同程度地发生了疖状腐蚀.用扫描电镜研究了氧化膜的显微组织.提出Zr-4合金发生疖状腐蚀的机制:Zr-4合金腐蚀生成的部分氧化膜具有微孔和微裂纹少、比较致密的特性,生长到一定程度后,在应力作用下,局部薄弱区发生平行于O/M界面的开裂并不断扩大,造成表层氧化膜破裂,腐蚀介质水进入裂纹中,形成有效的供氧源,使局部腐蚀加速,发生不均匀腐蚀,这种不均匀腐蚀在适当条件下发展成疖状腐蚀.氧化膜局部产生了可向O/M界面提供充足氧的直接供氧源,是引发锆合金产生疖状腐蚀的最密切因素.所有与发生疖状腐蚀有关的其它因素,如合金元素、第二相的大小和分布、氧化膜生长各向异性等,都是通过对氧化膜相关性质的影响而发生作用.     

表面状态对Zr—4合金抗疖状腐蚀性能的影响

对不同表面状态的Ｚｒ－４管材在５００℃,１０．３ＭＰａ下的过热蒸汽中进行了腐蚀试验。实验证明：表面处理状态对合金抗疖状腐蚀性能有较大影响。经酸洗后可提高抗疖状腐蚀性能,预生氧化膜处理可起到延缓疖状腐蚀的作用。结果表明,酸洗后预膜处理的表面有最好的抗疖状腐蚀性能。     

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.     

Zirconium alloys in nuclear power engineering

Results of studies of zirconium alloys É110 and É635 that have served in parts of VVÉR-1000 reactors are presented. The influence of the composition on the properties of alloys É110 and É635 is studied and improved modifications are suggested. The effect of the total content of admixtures in alloy É110 on corrosion and embrittlement of pipes under conditions simulating LOCA is investigated.     

Uniform and “Nodule” Corrosion of Zirconium Alloys under Service Conditions

Results concerning uniform and nodule (local) corrosion obtained in SF NIKIÉT are reviewed. The applicability of the electrotechnical theory of high-temperature oxidation of metals to zirconium alloys is analyzed. The conditions of occurrence of nodule corrosion are determined, and the results of a study of process channels and fuel assemblies of RBMK reactors after service at nuclear power plants (NPP) are generalized.     

Zirconium alloys in nuclear power engineering

Results of studies of zirconium alloys É110 and É635 that have served in parts of VVÉR-1000 reactors are presented. The influence of the composition on the properties of alloys É110 and É635 is studied and improved modifications are suggested. The effect of the total content of admixtures in alloy É110 on corrosion and embrittlement of pipes under conditions simulating LOCA is investigated.     

P2缓蚀剂处理铝合金的耐腐蚀性能研究

目的：通过加入一种合成缓蚀剂提高锆化液防锈性能。方法将P2缓蚀剂磷酸三乙醇胺加入到锆化液中处理铝合金,使其表面形成一层新型锆化膜。用扫描电镜(SEM)分析处理后样品放大20000倍的表面,并与原液处理的样品进行对比,通过能谱(EDS)及XRD分析微团成分,辅助使用硫酸铜点滴试验判断耐腐蚀能力的增强程度,并用百格试验测试处理后表面对漆膜的附着力变化,通过酸性盐雾试验(CASS)分析新型锆化膜处理后的附着力变化程度。结果使用加入P2缓蚀剂的锆化液的辅助硫酸铜点滴实验,较原样颜色变化所用的时间提高了48 s,用加入缓蚀剂的锆化液处理的样品中,氧元素较原样品中氧元素减少了一半,判断微团为氧化微团且在加入缓蚀剂的锆化液处理后大量减少,百格实验效果相当,能通过480 h的CASS。结论使用添加P2缓蚀剂的锆化液处理铝合金,能增强其耐腐蚀性能。     

LWR pellet-cladding interactions: Materials solutions to SCC

Zirconium alloys are commonly used as fuel-cladding tubes in water reactors because of their inherent resistance to a variety of environmental conditions. One of the major fuel-reliability issues of the 1970s and early 1980s was pellet cladding interaction (PCI). The mechanism of PCI is one of stress corrosion cracking (SCC) by a combination of aggressive fission products and cladding stress from pellet expansion. The severity of the problem, in particular in boiling water reactors, led to the development of barrier cladding by co-extrusion of Zircaloy-2 with an inner iodide zirconium that essentially eliminated the PCI-related failures. However, the substantially lower corrosion resistance of the zirconium layer led to clad breach and failures by other mechanisms. The difference in corrosion resistance could lead to some dramatic differences in post-failure fuel operations. This article briefly summarizes how PCI-SCC factors led to the development of PCI-resistant fuel cladding and concludes with a note on future research needs. For more information, contact K.L. Murty, North Carolina State University, Department of Nuclear Engineering, Raleigh, NC 27695-7909 USA; (919) 515-3657; fax (919) 515-5115; e-mail murty@eos.ncsu.edu.     

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。     

Ein neuer Weg zur Verbesserung der Korrosionsschutzschichten auf Stahl in Heißwasser

A new way to improve corrosion protection layers on steel exposed to hot water The corrosion of steels in hot water is self-stifled by the formation of oxide films. In cooling circuits of pressurized water reactors the residual corrosion is still high enough to produce radioactive corrosion products to such a degree as to complicate operations. A further lowering of the corrosion rate demands a more effective inhibition of the transport of corrosion reactants through the film. A closer investigation of the transport paths discussed so far showed that diffusion through water-filled pores dominates. It is postulated that the whole oxide layer forms by means of a combined mechanism of solution and deposition whereby pores are defined as interstices between the growing crystals. On the whole, an increase in the corrosion-protective value of the film is linked with a decrease in its porosity. This is attainable if, in the initial phase of the film growth, the corrosive environment contains titanium or zirconium compounds originating hydrous oxides fit for deposition upon the materials surface in subcolloidal dispersion. The resulting hindrance of the growth of oxide crystals leads to the formation of thin, apparently amorphous films having improved corrosion-protective properties.     

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.