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.     

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO₂–zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10–22 Cr + 4–6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.     

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.     

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.     

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。     

High-density U-Mo fuel for research reactors

The effect of the content of the main alloying element (molybdenum in an amount of from 1.5 to 9 wt.%) and of additives of aluminum, silicon, and tin on the structure, phase composition, and properties of uranium alloys used as dispersion nuclear fuel in research reactors is studied. The heat treatment parameters ensuring a stable structure in the alloys are determined. Segregation of secondary phases in uranium alloys with molybdenum and molybdenum and tin additives is studied. Heat treatment parameters optimum for providing fine segregations of UAl₂ and U₅Sn₄ particles in the matrix are determined. The results of the study are used to develop a composition for fabricating fuel grit and dispersion fuel for reactor tests.     

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.     

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.     

Phase Characteristics of a U-22Pu-4Am-2Np-40Zr Metallic Alloy Containing Rare Earths

Metallic fuel alloys consisting of uranium (U), plutonium (Pu), and zirconium (Zr) with minor additions of americium (Am) and neptunium (Np) are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. The current irradiation test series design, designated Advanced Fuel Cycle-2 (AFC2), includes minor additions of rare earth (RE) elements to simulate expected fission product carryover from the electrochemical molten salt reprocessing technique. The as-cast fuel alloys have been investigated for phase and thermal properties; specifically, enthalpies of transition, transition temperatures, and room temperature phase characteristics. Results and observations related to these characteristics for the “fresh” fuel alloys are provided. The alloy compositions are based on a U-22Pu-4Am-2Np-40Zr alloy, along with additions of 1.3 and 1.9 at.% RE (at the expense of uranium where RE denotes rare earth alloy of cerium, lanthanum, praseodymium, and neodymium). Phase behavior and associated transitions have been compared to available U-Pu-Zr ternary diagrams with acceptable agreement. Enthalpies of transition were deconvoluted from heating and cooling thermal traces for relatively reliable values. The RE additions to the base alloy have a minimal influence on the room temperature phases present and phase transition temperatures, but the room temperature phases present did impact the enthalpies of transition.     

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。     

Hydrogen Charging of Zirconium Parts in the Process of Manufacturing and Operation: a Factor Limiting Their Life in VVÉR and RBMK Reactors

The causes of hydrogen charging of zirconium parts in cores of VVÉR and RBMK reactors during manufacturing and operation and the factors influencing these processes are considered. The laws governing the influence of hydrogen charging on the fracture toughness of thin-wall zirconium parts are studied. It is shown that cracks may develop in thin-wall zirconium parts of fuel rod arrays under the action of hydrogen by the mechanism of delayed hydride cracking. The effect of irradiation on the rate of crack propagation in zirconium parts due to delayed hydride cracking is considered.     

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.     

How the crystallography and nanoscale chemistry of the metal/oxide interface develops during the aqueous oxidation of zirconium cladding alloys

Aqueous corrosion and hydrogenation have become major limiting factors to the use of zirconium alloys as fuel cladding and assembly components in water-cooled nuclear reactors. The metal–oxide interface has been a particular focus of previous research, but there is still no clear understanding of what is present at the interface at different stages of the complex oxidation process. We report here a systematic investigation using state-of-the-art instrumentation on the interfaces in several zirconium alloys corroded for different times. We have shown that thin intermediate oxide layers with compositions close to ZrO can be observed in almost all the pre-transition samples studied, and that this layer thickens during the pre-transition stage. Just before the kinetic transition, a large variation in the suboxide width was detected, suggesting that the kinetic transition is an extremely local process. After transition the suboxide was generally absent. In the suboxide locations different structures, including an unidentified phase, were found. The oxygen-saturated (～30 at.% O) metal regions found beneath the oxide are thickest in the (late) pre-transition samples and significantly thinner in the post-transition samples. We suggest that the suboxide cannot by itself act as a protective layer and conclude that it is the development of interlinked porosity down to the metal–oxide interface that is the reason for the transition in oxidation kinetics.     

A Theoretical Study of the Kinetics of Hydride Cracking in Zirconium Alloys

A diffusion model is suggested for computing the rate of delayed hydride cracking (DHC) in nonirradiated zirconium-base alloys. The rates of DHC in claddings of fuel elements of VVÉR and RBMK reactors and in pressure pipes of CANDU reactors are predicted.     

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,其提升程度存在一定的差别。位错滑移是锆及锆合金高周疲劳的主要损伤机制,位错运动(包括位错滑移及位错攀移)是锆及锆合金低周疲劳的主要损伤机制。文章最后指出现阶段锆及锆合金SPD技术的发展趋势及应用前景。     

Potentialities of Mössbauer Spectroscopy for Studying Zirconium Alloys and Their Oxide Films

A brief review and an analysis of results of a study of iron- and tin-bearing zirconium alloys and their oxide films by the method of Mössbauer spectroscopy (MS) are presented. The potentialities of MS for studying the phase composition of zirconium alloys are described and the changes in the states of iron and tin atoms are presented as a function of additional alloying and thermomechanical treatment. The conditions of formation of Zr₂Fe and Zr₃Fe intermetallic compounds and chromium- and niobium-bearing compounds are considered. It is shown that some intermetallic compounds transform into other compounds at room temperature. Metallic iron and tin are shown to be present in oxide films of zirconium alloys, and their concentration is shown to affect the corrosion resistance of zirconium alloys.     

含Nb锆合金第二相及其与腐蚀行为关系研究进展

含Nb锆合金具有优异的耐腐蚀性能、良好的机械性能和加工变形能力,是目前锆合金研究的重点。本文综述了近年来含Nb锆合金的研究现状,包括化学成分、变形及热处理工艺对第二相粒子析出的影响,介绍了锆合金腐蚀理论的研究进展,讨论了研究中存在的若干问题,为含Nb锆合金组织控制和耐腐蚀性能改善提供参考。     

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.     

Kinetics of high-temperature oxidation and embrittlement factors of zirconium alloys in tests simulating LOCA-type accidents in nuclear power plants

Tests of tube samples of fuel claddings of zirconium alloys of different chemical composition were performed with simulation of emergency situations at nuclear power plants with coolant loss including heating, high-temperature oxidation in steam, and cooling. Kinetics of high-temperature oxidation and structural phase changes under heating and cooling were studied, structure and analysis of fractures were studied quantitatively, and mechanic properties (residual ductility) of the samples after cooling were determined. The principal causes of alloy embrittlement were determined. Irrespective of the observed changes in the oxidation kinetics and microstructure character, the residual ductility of all the studied alloys was zero.     

The development of metallic nuclear fuels for transmutation applications: Materials challenges

Metallic alloys show great potential to serve as transmutation fuels that could be used to burn long-lived and high-heat-producing minor actinides and fission products in nuclear reactors as part of the Global Nuclear Energy Partnership program. To implement these fuels, work is ongoing to develop fuel fabrication processes, characterize alloy microstructures, measure fuel properties, determine the compatibility of fuel and cladding alloys, and understand the performance of the fuel alloys during irradiation.