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.     

Development of high strength zirconium alloys

Abstract

Assessment of the effects of various alloying elements and strengthening methods in zirconium alloys has indicated that alloys with adequate strengtl1 for use as pressure tube material at temperatures above 400°C can be developed. Ternary, quaternary and quinary alloys based on zirconium have been prepared and subjected to various thermal and deformation treatments. Their age-hardening behaviour and tensile properties are described and some information on their corrosion resistance and response to fast neutron irradiation at 300° and 450°C is presented. The target of UTS in the range 90–120 kpsi at 450°C can be achieved in alloys having an acceptable thermal neutron capture cross section. Adequate strength combined with the ductility necessary for pressure tube fabrication can be obtained by air cooling selected alloys from temperatures high in the (α + β) range.

Résumé

L'evaluation des effets de divers éléments d'alliage et des méthodes dc durcissement permet de croire que des alliagcs ayant une résistance adéquate peuvent être développés pour des tubes de force utilisés à des températures supérieures à 400°C. Des alliages ternaires, quaternaires et quénaires à base de zirconium ont été préparés et soumis à divers cycles thermiques et cycles de déformation. L'étude a porté sur leur vieillissement et leurs propriétés en traction; quelques observations sur leur tenue en corrosion et en irradiation par des neutrons rapides à 300° et 450°C sont présentées. Il est possible d'atteindre des résistances à la traction de 90 à 120 kpsi a 450°C pour des alliages ayant une faible section de capture des neutrons. Une résistance suffisante jointe à la ductilité nécessaire pour la fabrication des tubes peut être obtenue par refroidissement à l'air de certains alliages à partir d'une température élevée dans le domaine α + β.     

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_丁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。     

Hot Working of Zirconium Alloys: Some Recent Developments

Abstract

The published results on the hot deformation behaviour of zirconium and its alloys have been reviewed critically to ascertain the type of information available in the areas of hot workability and evolution of microstructure during hot deformation. It is observed that in a majority of these investigations, the emphasis was on the fundamental aspects of hot deformation such as identification of the rate controlling mechanisms, activation energy for hot working and identification of softening mechanisms. However, no systematic effort has been directed towards the optimization of hot workability and control of microstructure of these technologically important materials. In recent years efforts have been made to characterize the hot workability and development of microstructure during hot deformation by constructing power dissipation maps or processing maps and flow instability maps in the format of Dynamic Materials Model. The power dissipation maps and instability maps have been found very useful in optimizing the hot workability and controlling the microstructure during hot deformation of several zirconium alloys.     

Corrosion of Refractory Alloys in Molten Lithium and Lithium Chloride

Abstract

Several refractory materials have been considered over the years as containment material for lithium and lithium halides. Surface modified refractory metals are being extensively investigated for containment of reactive metals, radionuclides and their compounds. An overview of experimental observations and results of liquid lithium corrosion of selected engineering refractory materials are presented. The nature of the degradation and its mechanism has been explained. The influence of temperature, microstructure, stress, impurities and service time on the corrosion behavior for various refractory alloys have been discussed. Selection rules for materials of containment for liquid lithium and lithium compounds have been suggested. Recent experimental observations on the behavior of tantalum and niobium-based refractory metal alloys in a specific molten salt environment comprising LiCl/Li₂O/Li/Li₃N at 725^°C have been included in an effort to select suitable materials for molten salt equipment. It has been observed that oxygen contamination is particularly harmful for the refractory metal alloys where as nitrogen is deleterious to iron-based alloys.     

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。     

Processing of Secondary Sources of Refractory Metals

Abstract

Refractory metals belonging to the Group IV to VIA of the periodic table have assumed very important role in variety of high-tech applications. Extraction of these metals so far has been accomplished mainly from primary and byproduct sources. Depletion of primary sources, non-availability of the sources of these metals in certain countries, relatively large applications and greater emphasis on environmental preservation have led to increased awareness in using secondary sources of these metals. Use of secondary sources in fact makes a better economic source particularly for these high value metals

This paper is based on extensive studies carried out in the Materials Group of BARC on the recovery of refractory metals such as tungsten, molybdenum, niobium, tantalum, vanadium and zirconium from a variety of secondary sources available within the country. Major objective has been to develop processes which can be operated even on a small scale with variable grade of the source using indigenously available equipment     

Effects of transient carbon deposition on the sodium sulfate-induced corrosion of nickel-base alloys

The role of solid carbon in accelerating the Na₂SO₄-induced corrosion of nickel-base alloys at high temperature has been investigated. Corrosion rates of salt-coated IN 738 alloy in an oxidizing atmosphere containing 0.1 pct SO₂ were much increased when the alloy was initially exposed to a low concentration of methane for a short time at 1273 K (1000°C). Crucible tests carried out with Ni-5 Al, Ni-30 Al, Ni-20Cr-15 Al and IN 738 specimens immersed in Na₂SO₄-carbon mixtures showed that corrosion of these alloys was dramatically enhanced if the carbon persisted long enough to form a Na₂S phase by reaction with the Na₂SO₄. In an oxidizing environment a graphitic carbon had a more pernicious effect on alloy corrosion resistance than a finely dispersed carbon black which burnt away more readily. Corrosion rates in air were also accelerated in the presence of carbonized crude oil residue such as might be formed as a result of incomplete fuel combustion in a gas turbine.     

Study on Alloying Behaviour of Uranium Zirconium Alloy

Uranium–zirconium alloy is the potential candidate material as metallic fuels for nuclear reactor applications. Fabrication of uranium zirconium alloy can be made either by powder metallurgy or by melting method. Both the method has its own advantages and selection of the fabrication route depends on its application as fuel. In this work investigations were carried out on fabrication of uranium–zirconium alloys (i.e. U–40 wt% Zr, U–50 wt% Zr, U–60 wt% Zr and U–70 wt% Zr) using both powder metallurgy and melting method. It is heat treated at different conditions. Apart from other characterization, X-ray diffraction analysis was carried out to evaluate phase content of the alloy and reported here.     

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。     

Zirconium alloys dispersion strengthened with yttria

Abstract

A powder metallurgy process for the manufacture of zirconium alloys dispersion strengthened with yttria particles is described. The resulting alloys con tain a suitable dispersion of stable yttria particles in a fine-grained matrix. Tensile and stress rupture tests over a range of temperature on Zircaloy-2 containing 0, 5 and 10 vol % yttria show the strength of yttria-containing alloys to be significantly improved. The post-irradiation tensile properties of Zircaloy-2 - yttria alloys have been determined with alloys irradiated at 280°C to 9.2 × 10²⁰ n/cm² thermal and 2.3 × 10²⁰ n/cm² fast. The alloys have been subjected to accelerated corrosion tests in steam and exhibit a variable response. Some possible detrimental factors and means of improvemen t of properties are discussed.

The mechanical properties of Zircaloy-2 alloys containing yttria particles are compared to some other zirconium alloys, Zircaloy-2 and zirconium-2.5 wt % Nb and the comparisons show that the dispersion- strengthened alloys arc worthy of further development.

Résumé

La fabrication par frittage des alliages de zirconium renforcés par des particules d'yttria est décrite. Ce procédé donne une dispersion convenable des particules d'yttria dans une matrice à grains fins. Des essais de traction et d'impact, effectués à diverses températures sur du Zircaloy-2 contenant 0,5 et 10 % d'yttria, ont montré que la résistance mécanique des alliages est nettement augmentée. D.autres essais de traction ont été effectués sur des alliages Zircaloy-2-yttria aprés des irradiations de 9.2 × 10²⁰ n/cm² thermique et 2.3 × 10²⁰ n/cm² rapide à 280°C. Les alliages ont été soumis à des essais de corrosion accélérée (vapeur) avec des résultats variables. Dans la discussion on considère les facteurs qui peuvent être miscibles ainsi que déautres améliorations possibles.

La comparaison des propriétés mécaniques des alliages de Zircaloy-2 contenant de l'yttria à celles d'autres alliages de zirconium, Zircaloy-2 et zirconium-2.5,% Nb indique que les alliages à particules dispersées méritent d'être étudiés davantage.     

Investigation of the processes of obtaining plastic treatment and electrochemical behaviour of lead alloys in their capacity as anodes during the electroextraction of zinc I. Behaviour of Pb---Ag, Pb---Ca and PB---Ag---Ca alloys

The influence of the mode of obtaining and plastic treatment of binary Pb---Ag and Pb---Ca, and ternary Pb---Ag---Ca alloys (used as anodes in Zn electro-extraction from sulphate electrolytes) on their electrochemical behaviour and corrosion resistance is studied.It has been established that the rolled Pb---Ag alloys possess a higher corrosion resistance and lower anodic polarization compared to the cast lead-silver ones due to the structural fineness and homogeneity of the plastic deformed anodes. The plastic deformation of the ternary alloys with calcium content of 0.06% causes Pb₃Ca precipitation in the solid solution. The hot-rolled alloys form an solid solution of Pb₃Ca with fine-grained structure, deformed through the rolling direction. The cold-rolled alloys possess clearly expressed oriented structures also through the rolling direction.Cast and plastically deformed Pb---Ca anodes possess better electrochemical and corrosion characteristics than pure Pb but a considerably higher anodic polarization and lower corrosion resistance than Pb---Ag alloys. Both Pb---Ca systems, with a calcium content of 0.08 and 0.11 % are very appropriate for the preparation of ternary Pb---Ag---Ca alloys. These alloys (Pb---Ag 0.5 %-Ca 0.11 %) possess better electrochemical and corrosion characteristics than binary Pb---Ca ones. Only ternary rolled alloys have equal corrosion and electrochemical properties to those of the alloy Pb---Ag (0.75%–1.0%) used in practice.     

PREPARATION OF ZIRCONIUM-COPPER AND ZIRCONIUM-COPPER-MOLYBDENUM ALLOYS BY SINTERING THE HYDRIDE POWDERS

Abstract

The results are reported of a laboratory investigation aimed at developing a method of recovering machining scrap from zirconium alloys by powder-metallurgy methods. Observations made during the crushing, and sintering of the hydride are described. The mechanical properties obtained before and after heat-treatment are reviewed, together with the resistance to corrosion by carbon dioxide of these alloys at 700°C. In general, the properties of the dense sintered parts arecomparable with those of cast parts, quite apart from the hardening influence of the oxygen content.     

Investigation of the processes of obtaining plastic treatment and electrochemical behaviour of lead alloys in their capacity as anodes during the electroextraction of zinc I. Behaviour of PbAg, PbCa and PBAgCa alloys

The influence of the mode of obtaining and plastic treatment of binary PbAg and PbCa, and ternary PbAgCa alloys (used as anodes in Zn electro-extraction from sulphate electrolytes) on their electrochemical behaviour and corrosion resistance is studied.It has been established that the rolled PbAg alloys possess a higher corrosion resistance and lower anodic polarization compared to the cast lead-silver ones due to the structural fineness and homogeneity of the plastic deformed anodes. The plastic deformation of the ternary alloys with calcium content of 0.06% causes Pb₃Ca precipitation in the solid solution. The hot-rolled alloys form an solid solution of Pb₃Ca with fine-grained structure, deformed through the rolling direction. The cold-rolled alloys possess clearly expressed oriented structures also through the rolling direction.Cast and plastically deformed PbCa anodes possess better electrochemical and corrosion characteristics than pure Pb but a considerably higher anodic polarization and lower corrosion resistance than PbAg alloys. Both PbCa systems, with a calcium content of 0.08 and 0.11 % are very appropriate for the preparation of ternary PbAgCa alloys. These alloys (PbAg 0.5 %-Ca 0.11 %) possess better electrochemical and corrosion characteristics than binary PbCa ones. Only ternary rolled alloys have equal corrosion and electrochemical properties to those of the alloy PbAg (0.75%–1.0%) used in practice.     

Development of single-phased copper alloy for seawater applications: As a cost-effective substitute for Cu-10Ni alloy

Cu-10Ni alloy is a standard material for seawater application. Under Indian scenario, where half of copper and whole of Ni (and Cu-Ni) requirement is met by imports, search for a cheaper substitute is obvious. An attempt has been made to develop a single-phased copper alloy containing 10 wt% Ni, 29 wt% Zn, up to 5 wt% Mn and 1 wt% Fe as a substitute to Cu-10wt% Ni-1wt% Fe alloy. Studies on Corrosion resistance of test alloys were carried out in synthetic seawater (ASTM D 114-75) by electro-chemical methods (cathodic and anodic polarization). Corrosion product film formed on alloy containing 5wt% Mn was characterized by SEM and XRD. A film of Cu₂O was found to form on the surface of the alloy, which accords corrosion protection. The test alloys containing 29 wt% Zn and 3 or 5 wt% Mn have exhibited better corrosion resistance in synthetic seawater than Cu-10Ni alloy. The role of Mn and Zn addition has been discussed.     

Corrosion resistance of new copper alloy containing 29Zn, 10Ni and up to 5Mn vis-a-vis Cu-10Ni in sulphide polluted synthetic seawater

Cu-10 Ni alloy suffers from accelerated corrosion in sulphide-polluted seawater. New copper alloys containing 29% Zn, 10Ni and 3 or 5% Mn have been found to be more corrosion resistant than Cu-10Ni alloys in sulphide polluted synthetic seawater. The studies were carried out in synthetic seawater (ASTM D 114–75) and sulphide was added through Na₂S (1 gpl ≈ 41ppm S²⁻). Testing was carried out using cathodic and anodic potentiodynamic polarization methods. Corrosion product film formed on the Cu-29Zn-10Ni-5Mn alloy in sulphide polluted synthetic seawater was characterized using SEM and XRD. The film was found to contain Cu₂O, Cu₂S, NiS and ZnS. Better corrosion resistance of new alloys as compared to that of Cu-10Ni alloy is attributed to formation of ZnS, a bad conductor, in the film and incorporation of Mn³⁺ cations in Cu₂O and Cu₂S lattice.     

A Study on Zirconium Rich Uranium–Zirconium Alloys

Uranium–zirconium alloy is the potential candidate materials as metallic fuels for nuclear reactor applications. Along with other properties, volumetric stability is an important issue for these types of fuels. In this work investigations were carried out on the zirconium rich uranium–zirconium alloys (i.e. U–50 wt% Zr, U–60 wt% Zr and U–70 wt% Zr) in as-cast as well as in heat treated conditions. Microstructural and dilatometric studies were carried out along with X-ray diffraction analysis to evaluate the phase content as well as the phase transformation behaviour of these alloys under different heat treated conditions.     

Microstructure and phase identification in type 304 stainless steel-zirconium alloys

Stainless steel-zirconium alloys have been developed at Argonne National Laboratory to contain radioactive metal isotopes isolated from spent nuclear fuel. This article discusses the various phases that are formed in as-cast alloys of type 304 stainless steel and zirconium that contain up to 92 wt pct Zr. Microstructural characterization was performed by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS), and crystal structure information was obtained by X-ray diffraction. Type 304SS-Zr alloys with 5 and 10 wt pct Zr have a three-phase microstructure—austenite, ferrite, and the Laves intermetallic, Zr(Fe,Cr,Ni)_2+x. whereas alloys with 15, 20, and 30 wt pct Zr contain only two phases—ferrite and Zr(Fe,Cr,Ni)_2+x. Alloys with 45 to 67 wt pct Zr contain a mixture of Zr(Fe,Cr,Ni)_2+x and Zr₂(Ni,Fe), whereas alloys with 83 and 92 wt pct Zr contain three phases—α-Zr, Zr₂(Ni,Fe), and Zr(Fe,Cr,Ni)_2+x. Fe₃Zr-type and Zr₃Fe-type phases were not observed in the type 304SS-Zr alloys. The changes in alloy microstructure with zirconium content have been correlated to the Fe-Zr binary phase diagram.     

Thermodynamics of the oxygen solutions in zirconium-containing iron-nickel melts

Thermodynamic analysis of the oxygen solutions in zirconium-containing iron-nickel melts is carried out. The equilibrium deoxidation constants of the melts by zirconium, the activity coefficients at infinite dilution, and the interaction parameters in melts of various compositions are determined. The dependences of the oxygen solubility in the melts on the nickel or zirconium content are calculated. Zirconium is shown to possess a very high deoxidizing capacity in iron-nickel alloys. The zirconium contents at the minima in oxygen solubility curves and the corresponding minimum oxygen concentrations are determined. As the nickel content in a melt increases to ∼45%, the deoxidizing capacity of zirconium decreases and, then, increases. The deoxidizing capacity of zirconium in pure nickel is noticeably higher than that in pure iron.