Comprehensive nuclear materials 2 07 zirconium alloys properties and characteristics Comprehensive nuclear materials 2 07 zirconium alloys properties and characteristics Comprehensive nuclear materials 2 07 zirconium alloys properties and characteristics Comprehensive nuclear materials 2 07 zirconium alloys properties and characteristics Comprehensive nuclear materials 2 07 zirconium alloys properties and characteristics
2.07 Zirconium Alloys: Properties and Characteristics C Lemaignan Commissariat a` l’E´nergie Atomique, Grenoble, France ß 2012 Elsevier Ltd All rights reserved 2.07.1 Introduction 217 2.07.2 2.07.3 2.07.3.1 2.07.3.1.1 2.07.3.1.2 2.07.3.1.3 2.07.3.2 2.07.3.3 2.07.3.4 2.07.3.4.1 2.07.4 2.07.4.1 2.07.4.2 2.07.5 2.07.5.1 2.07.5.2 2.07.5.3 2.07.5.4 References Physical Properties Alloy Processing Nuclear Grade Zr Base Metal Ore decomposition Hf purification and removal Reduction to the metal Alloy Melting Forging Tube Processing Crystallographic texture development Alloys Alloying Elements and Phase Diagrams Industrial Alloys Mechanical Properties Before Irradiation Strength and Ductility Mechanical Properties in Temperature and Creep Hydrogen Embrittlement and Other H Effects Prospectives 218 219 219 219 219 220 220 220 221 221 222 222 227 228 228 229 229 230 231 Abbreviations ASTM American Society for Testing Materials BWR Boiling water reactor CANDU Canada Deuterium Uranium (heavy water pressurized reactor) DHC Delayed hydride cracking DSA Dynamic Strain Aging hcp Hexagonal closed packed HPUF Hydrogen pick-up fraction LOCA Loss of coolant accident MIBK Methyl-isobutyl-ketone PWR Pressurized water reactor RBMK Reaktor Bolshoy Moshchnosti Kanalniy (pressure tube power reactor) RIA Reactivity induced accident SPP Second phase particle TEM Transmission Electron Microscope TM Transition metal VVER Voda-Voda Energy Reactor (PWR of Russian design) 2.07.1 Introduction Zirconium (Zr) exhibits a physical property of uppermost importance with respect to the design of in-core components of thermal neutron power reactors: it has a very low thermal neutron capture cross-section, and its alloys exhibit good engineering properties For an improvement in neutron efficiency of the water-cooled reactors, the development of industrial-type Zr-based alloys started as early as the beginning of the nuclear reactor design, and is still continuing The engineering properties of Zr and Zr alloys are therefore widely studied Information exchanges and reviews are available in various sources; for example, the International Atomic Energy Agency issued reviews on Zr alloys for nuclear applications For more detailed, up-to-date information, the reader is referred to a recent one,1 or to the proceedings of the symposia on ‘Zr in the nuclear industry,’ organized at 2–3 year intervals by ASTM.2 217 218 Zirconium Alloys: Properties and Characteristics It was found early that Zr is naturally mixed in its ore with its lower companion of the periodic table, hafnium, the latter being a strong neutron absorber Purification of Zr from Hf contamination is therefore mandatory for nuclear applications The development of the industrial alloys has been performed following the classical route: searching for elements of significant solubility that would improve the engineering properties, without too much impact on the nuclear ones Tin, niobium, and oxygen are the main alloying elements, with minor additions of transition metals (TMs) (Fe, Cr, and Ni) Heat treatments aiming at homogeneous solid solutions, phase transformations, and precipitation control allow optimizing the structure of the alloys In addition, the thermomechanical history of the components strongly impacts their behavior, via the formation of a crystallographic texture, because of the anisotropy linked to the hexagonal crystallography of Zr at low temperature A few Zr alloys are commonly used for structural components and fuel cladding in thermal neutron reactors Zircaloy (Zry)-4 is used in pressurized water reactors (PWRs) and Zircaloy-2 in boiling water reactors (BWRs) The heavy water-moderated CANDU reactors, as well as the Russian VVER or RBMK reactors, use Zr–Nb alloys New alloys are designed based on variants of the Zr–1% Nb, with small additions of Fe and sharp control of minor additions (M5®), or variants of the quaternary alloys, such as Zirlo® and E635 More complex alloys with other types of alloying elements are also being tested in power plants, but the actual experience accumulated on these alloys is too low to consider them as commonly accepted, from an industrial point of view Fuel claddings are made out of Zry-2 or Zry-4 Those tubes have different geometries, depending on reactor design In PWR’s, the fuel cladding rods are 4–5 m long and have a diameter of 9–12 mm for a thickness of 0.6–0.8 mm BWR fuel rods are usually slightly larger The design is similar for the Russian VVER, with Zr–1%Nb In CANDUs, the fuel bundles are shorter (0.5 m) and the cladding is thinner (0.4 mm) in order to collapse very fast on the UO2 pellets Structural components of zirconium alloys are the guide tubes, the grids, and the end plates that maintain the components of the fuel assemblies They have to maintain the structural integrity at the stress levels corresponding to normal or accidental operations In addition, they should have very low corrosion rates in the hot, oxidizing coolant water In BWRs, each assembly is surrounded by a Zircaloy-2 channel box that avoids cross-flow instabilities of the two-phase coolant Their geometrical stability is a mandatory requirement for the neutron physics design of the core In the case of CANDUs and RBMKs, the moderator is separated from the coolant water The coolant water in contact with the fuel rods is contained in pressure tubes, usually made of Zr–Nb alloys They are large components (L $ 10 m, [ $ 30 cm, and e $ mm), with a design life expected to match the reactor life, that is, tens of years, with only minor corrosion and creep deformation 2.07.2 Physical Properties Natural zirconium has an atomic mass of 91.22 amu, with five stable isotopes (90Zr : 51.46%, 91Zr: 11.23%, 92 Zr: 17.11%, 94Zr: 17.4%, and 96Zr: 2.8%) The depletion of the most absorbing isotope (91Zr, with sa $ 1.25 Â 10À28 m2) would increase further the interest of using Zr alloys in reactors, but would clearly be economically inefficient The cross-section for elastic interaction with neutrons is normal, with respect to its atomic number (sdiff $ 6.5 barn) Despite its high atomic mass, the large interatomic distance in the hcp crystals lead to a limited specific mass of 6.5 kg dmÀ3 The thermophysical properties correspond to standard metals: thermal conductivity $22 W mÀ1 KÀ1 and heat capacity $280 J kgÀ1 KÀ1, that is, close to 3R per mole Below 865 C, pure Zr has an hcp structure, with a c/a ratio of 1.593 (slightly lower than the ideal 1.633) The lattice parameters are a ¼ 0.323 nm and c ¼ 0.515 nm.3 The thermal expansion coefficients show a strong anisotropy, with almost a twofold difference between the aa and ac coefficients (respectively 5.2 and 10.4 Â 10À6 KÀ1).4 This anisotropic behavior of the thermal expansion induces internal stresses due to strain incompatibilities: After a standard heat treatment of 500 C, where the residual stresses will relax, cooling down to room temperature will result in internal stresses in the range of 100 MPa, depending on grain-to-grain orientations The modulus of elasticity is also anisotropic, but with lower differences than for thermal expansion (Ea ¼ 99 GPa, and Ec ¼ 125 GPa).5 For industrial parts, the values recommended are close to a $ 6.5 Â 10À6 KÀ1 and E $ 96 GPa The temperature evolution of the elasticity constants is unusual: the elasticity is strongly reduced as the temperature increases ($5% per 100 K).6,7 This abnormal behavior is specific to the hcp metals of the IV-B row of the periodic table.8 Zirconium Alloys: Properties and Characteristics 219 tons per year, out of which only 5% is processed into zirconium metal and alloys The processing of Zr alloy industrial components is rather difficult because of the high reactivity of the Zr metal with oxygen It consists of several steps to obtain the Hf-free Zr base metal for alloy preparation: decomposition of the ore to separate Zr and Si, Hf purification, and Zr chloride or fluoride reduction 50 mm 2.07.3.1.1 Ore decomposition Three different processes are currently used for the Zr–Si separation: Figure Microstructure of a b-quenched Zr alloys, with a-platelets of four different crystallographic orientations issued from the same former b-grain At 865 C, Zr undergoes an allotropic transformation from the low temperature hcp a-phase to the bcc b-phase On cooling, the transformation is usually bainitic, but martensitic transformation is obtained for very high cooling rates (above 500 K sÀ1) The bainitic transformation occurs according to the epitaxy of the a-platelets on the old b-grains, as proposed by Burgers9,10: (0001)a // {110}b and h1120ia // h111ib Among the 12 different possible variant orientations of the new a-grains, only a few are nucleated out of a given former b-grain during this transformation to minimize the internal elastic strain energy This process leads to a typical ‘basket-weave’ microstructure (Figure 1) As a result, a b-quenching does not completely clear out the initial crystallographic texture that had been induced by the former thermomechanical processing.11,12 Although the alloying elements present in the Zr alloys change the transformation temperatures, with a 150 C temperature domain in which the a- and b-phases coexist, the crystallographic nature of the a-b transformation is equivalent to that of pure Zr Specific chemical considerations (segregations and precipitations) will be described later The melting of pure Zr occurs at 1860 C, significantly above the melting temperature of other structural alloys, such as the structural or stainless steels At high pressures, (P > 2.2 GPa) a low-density hexagonal structure is observed, known as the o-phase 2.07.3 Alloy Processing 2.07.3.1 Nuclear Grade Zr Base Metal The most frequently used ore is zircon (ZrSiO4), with a worldwide production of about million metric In alkali fusion, where the zircon is molten in a NaOH bath at 600 C, the following reaction takes place: ZrSiO4 ỵ 4NaOH ! Na2 ZrO3 ỵ Na2 SiO3 ỵ 2H2 O Water or acid leaching allows the precipitation of ZrO2 The fluo-silicate fusion: ZrSiO4 ỵ K2 SiF6 ! K2 ZrF6 ỵ 2SiO2 It produces a potassium hexafluorozirconate which, reacting with ammonia, leads to Zr hydroxide The carbo-chlorination process is performed in a fluidized bed furnace at 1200 C The reaction scheme is the following: ZrO2 ỵSiO2 ỵ HfO2 ị ỵ 2C ỵ 2Cl2 ! ZrCl4 ỵSiCl4 ỵ HfCl4 ị ỵ 2C The controlled condensation of the gaseous tetrachloride allows the separation of Zr and Si, but not of Hf from Zr 2.07.3.1.2 Hf purification and removal The processes described above separate Si from Zr, but the Zr compounds remain contaminated with the initial Hf concentration The high neutron capture cross-section of Hf (sa $ 105 barn, compared to 0.185 barn for Zr) requires its suppression in Zr alloys for nuclear application Two major processes are used for this step: the MIBK-thiocyanate solvent extraction and the extractive distillation of tetrachlorides In the first case, after reaction of zirconyl chloride (ZrOCl2), obtained by hydrolysis of ZrCl4, with ammonium thiocyanate (SCN-NH4), a solution of hafnyl-zirconyl-thiocyanate (Zr/Hf )O(SCN)2 is obtained A liquid–liquid extraction is performed with methyl-isobutyl-ketone (MIBK, 220 Zirconium Alloys: Properties and Characteristics name of the process) Hf is extracted into the organic phase, while Zr remains in the aqueous one Hf-free ZrO2 is obtained after several other chemical steps: hydrochlorination, sulphation, neutralization with NH3, and calcination In the dry route, after the transformation of zircon into its chloride ZrCl4, through the carbochlorination process, Zr and Hf are separated using a vapor phase distillation, at 350 C, within a mixture of KCl-AlCl3, where the liquid phase is enriched in Zr, and the vapor in Hf 2.07.3.1.3 Reduction to the metal The final step to obtain metallic Zr of nuclear grade is to reduce the Hf-free Zr compounds that have been obtained by the previous steps Two processes are to be considered at an industrial scale: the Kroll process and the electrolysis In the Kroll process, the Zr metal is obtained by the reduction of ZrCl4 in gaseous form by liquid magnesium, at about 850 C in an oxygen-free environment The following reaction occurs: ZrCl4 gị ỵ 2Mglị ! MgCl2 lị ỵ Zrsị After distillation of the remaining Mg and MgCl2, under vacuum at 950 C, sintering of the Zr agglomerate at 1150 C gives the metallic sponge cake After wet chemical chemistry, the reduction of the ZrO2 obtained by the MIBK process is often performed by electrolysis It is realized with the mixed salt K2ZrF6 dissolved in NaCl or KCl at 850 C under inert gas, with stainless steel cathode on which Zr is deposited, and chlorine evolution at the graphite anode This route is mainly used in the Russian Federation, the names of the Russian alloys starting with an ‘E,’ referring to electrolytic processing High purity Zr can be obtained by the Van Arkel process It consists of reaction of Zr with iodine at moderate temperature, gaseous phase transport as ZrI4, and decomposition of the iodide at high temperature on an electrically heated filament The iodine released at the high temperature side is used for the low temperature reaction in a closed loop transport process, according to the following scheme: Zr + I2 => ZrI4 => Zr + I2 250–300 ЊC 1300–1400 ЊC This source of metallic Zr (called ‘iodide Zr’) is used in Russia in addition to Zr obtained by the electrolytic process for the melting of the alloys (typically 30% ‘iodide Zr’ in the first electrode to be melted) 2.07.3.2 Alloy Melting Whatever the processing route followed for the production of Zr metal, the sponge or the chips obtained by scrapping out the electrodes are the base products for alloy ingot preparation The melting of the alloys is performed using the vacuum arc remelting (VAR) process This process is specific to highly reactive metals such as Zr, Ti, or advanced superalloys For industrial alloy preparation, an electrode is prepared by compaction of pieces of base metal fragments (sponge or scraps) with inclusion of the alloying elements Typically, the elements to be added are the following: O (in the form of ZrO2 powder), Sn, Nb, Fe, Cr, and Ni to the desired composition In addition, a strict control of minor elements, such as C, N, S, and Si, is ensured by the producers, at concentrations in the range of 30–300 ppm, according to their requirements to fulfill the engineering properties A few specific impurities are strictly controlled for neutron physics reasons: Cd and Hf due to their impact on neutron capture cross-section, U for the contamination of the coolant by recoil fission fragments escaping from the free surface of the cladding, and Co for in-core activation, dissolution transport, contamination, and g-irradiation The compact stack is melted in a consumable electrode electric vacuum furnace with water chilled Cu crucible Electromagnetic fields are often used for efficient stirring of the liquid pool and reduced segregations After three to four melts, the typical dimensions of the final ingots are 0.6–0.8 m diameter and 2–3 m length, that is, a mass of 4–8 tons 2.07.3.3 Forging Industrial use of Zr alloys requires either tube- or plate-shaped material The first step in mechanical processing is forging or hot rolling in the b-phase, at a temperature near 1050 C, or at lower temperatures in the a ỵ b range or even in the upper a range The high oxidation kinetics of Zr alloys in air at high temperatures restricts the high temperature forging process to thick components, that is, with minimum dimensions larger than 10 cm, at least Final dimensions after forging correspond to 10–25 cm diameter for billets and 10 cm for slabs A b-quenching is usually performed at the end of the forging step This heat treatment allows complete Zirconium Alloys: Properties and Characteristics dissolution of the alloying elements in the b-phase and their homogenization above 1000 C, followed by a water quench During the corresponding bainitic b to a transformation, the alloying elements are redistributed, leading to local segregations: O and Sn preferring the middle of the a-platelets, while the TMs (Fe, Cr, and Ni) and Nb are being rejected to the interface between the platelets.13 These segregations lead to plastic deformation strains highly localized at the interplatelet zones for materials having a b-quenched structure (heat-affected zones, welds, or b-quenched without further thermomechanical processing) As described later, this b-quench controls the initial size distribution of the precipitates in Zircaloy, and further recovery heat treatments should be performed below the b–a transus only 2.07.3.4 Tube Processing For seamless tube production, first a hot extrusion is performed in the temperature range of 600–700 C For pressure tube fabrication, this step is followed by a single cold drawing step and a final stress relieving heat treatment For cladding tubes, the extrusion produces a large extruded tube (‘Trex’ or ‘shell’), of 50–80 mm in diameter and 15–20 mm in thickness, which is further reduced in size by cold rolling on pilger-rolling mills After each cold working step of plate or tube material, an annealing treatment is mandatory to restore ductility It is usually performed in the range of 530–600 C to obtain the fully recrystallized material (RX) The resultant microstructure is an equiaxed geometry of the Zr grains with the precipitates located at the a-grain boundaries or within the grains The location of the precipitates at the grain boundaries is not due to intergranular precipitation but because they pin the grain boundaries during grain growth (Figure 9) These different heat treatments contribute to the control of the cumulative annealing parameter to be described below For better mechanical properties of the final product, the temperature of the last annealing treatment can be reduced to avoid complete recrystallization This is the stress-relieved (SR) state, obtained with final heat treatment temperature of 475 C, which is characterized by elongated grains and a high density of dislocations, and by relief of the internal stresses, leading to a greater ductility than cold-worked materials It is mostly applied to the PWR claddings, while for BWRs, a complete recrystallization is performed at 550–570 C 221 2.07.3.4.1 Crystallographic texture development Two plastic deformation mechanisms are operating during low temperature deformation of the Zr alloys: dislocation slip and twinning As reviewed by Tenckhoff,14 the most active deformation mechanism depends on the relative orientation of the grain in the stress field Dislocation slip occurs mostly on prism plane with an a Burgers vector It is referred to as the {1010} h1210i, or prismatic, system The total strain imposed during mechanical processing of the Zr alloys cannot, however, be accounted for only with this single type of slip, as the different orientations of the crystal would only give two independent shear systems At high deformations, and as the temperature is increased, (c ỵ a) type slip is activated on {1121} or {1011} planes These are the pyramidal slip systems, having higher resolved shear stresses (Figure 2) Different twinning systems may be activated depending on the stress state: for tensile stress in the c-direction, {1012} h1011i twins are the most frequent, while the {1122} h1123i system is observed when compression is applied in the c-direction The resolved shear stresses of the twin systems have been shown to be higher than the one necessary for slip, but due to the dependence of the Schmid factor on orientation, twinning is activated before slip, for some well-oriented grains Therefore, there are five independent deformation mechanisms operating in each grain, and thus the von Mises criterion for grain-to-grain strain compatibility is fulfilled At the large strains obtained during mechanical processing, steady-state interactions occur between the twin and slip systems that tend to align the basal planes parallel to the direction of the main deformation.15,16 For cold-rolled materials (sheets or tubes), the textures are such that the majority of the Figure The two Burgers vectors (a and c ỵ a) for strain dislocations in Zr alloys, and the two slip planes (prismatic and pyramidal) in hcp a-Zr 222 Zirconium Alloys: Properties and Characteristics AD AD 90Њ 90Њ 80Њ 80Њ 60Њ 60Њ 30Њ 30Њ 5 1 0Њ TD 54 2 fR = 0.64 (a) 35 4Њ 0Њ TD fR = 0.55 (b) Figure h0001i Pole figure of two cladding tubes with slightly different mechanical processing routes grains have their c-axis tilted 30–40 away from the normal of the foil or of the tube surface toward the tangential direction, as can be seen in the h0001i pole figure of a cladding tube (Figure 3) During tube rolling, the spread of the texture can be reduced by action on the ratio of the thickness to diameter reductions (Q factor): a reduction in thickness higher than the reduction in diameter gives a more radial texture, that is, a texture with the c poles closer to the radial direction.16 After cold processing, the h1010i direction is parallel to the rolling direction, and during a recrystallization heat treatment a 30 rotation occurs around the c-direction and the rolling direction is then aligned with the h1120i direction for most of the grains 2.07.4 Alloys 2.07.4.1 Alloying Elements and Phase Diagrams Like any metal, pure Zr exhibits rather poor engineering properties To improve the properties of a given metal, the metallurgical engineering procedures are always the same: It consists in finding additions, any species of the periodic table could be considered, with significant solubility, or heat treatments producing new phases that could improve the properties The relative solubility of the various alloying elements in the a- and b-phases is therefore one basis for the choice of additions, as well as for developing the heat treatments, for microstructure control For the nuclear applications, neutron physics requirements restrict the possibilities, by rejection of the isotopes having high interaction cross-sections, or isotopes that would transmute to isotopes of high capture cross-section or having high irradiation impact (Co) Elements such as Hf, Cd, W, and Co have therefore not been considered for alloy developments With low nuclear impact, O, Sn, and Nb have been selected (Al and Si having also low nuclear impacts were not retained because of degradation in corrosion resistance), while other TMs (Fe, Cr, Ni, etc.) can be accepted up to limited concentrations (below 0.5% total) The additions have to improve the engineering properties The main properties to be improved are the corrosion behavior in hot water and the mechanical strength (yield stress, ductility, and creep) As described below, Sn and Nb are added for corrosion resistance, and elements forming secondary phases (Nb and Fe, Cr, and Ni) or solid solutions are also used for increasing the mechanical properties Last, the microstructure obtained after the thermomechanical processing should not change without control under irradiation Therefore, hardening obtained by precipitation or strain hardening can be considered only if the irradiation-induced evolution of the initial microstructure will be compensated by the development of irradiation-induced microstructural defects In this respect, the evolution of precipitates in Zircaloys is of high importance for corrosion behavior and geometrical integrity These points are discussed in Chapter 5.03, Corrosion of Zirconium Alloys and Chapter 4.01, Radiation Effects in Zirconium Alloys Most of the binary phase diagrams with Zr are already known and many ternary or higher-level Zirconium Alloys: Properties and Characteristics diagrams of industrial interest are now known.17 The need for a better control of the processing of the current alloys and the aim of finding new alloys and structures without too much experimental work have been a driving force for the modern trend in numerical simulation for material science It is now also possible to extrapolate the binary data to multicomponent systems In that respect, a thermodynamic database for Zr alloys, called ZIRCOBASE, has been developed under the Calphad methodology.18 This database contains 15 elements and is frequently updated The most complex ternary or quaternary phase diagrams available are optimized or computed using this database, and, in the case of missing basic thermodynamic data, with the contribution of ‘abinitio’ computations.19 The phase diagrams presented in this review were obtained according to this procedure Oxygen is highly soluble in the a-phase, and stabilizes at high temperature (Figure 4) Oxygen has to be considered as an alloying element This use of oxygen for strengthening is rare in metallurgy, compared to the use of nitrogen However, the use of nitrogen for strengthening would severely deteriorate the corrosion resistance, and nitrogen is removed as much as possible The purpose of oxygen additions is to increase the yield strength by solution strengthening, without degradation of the corrosion resistance The O content is not specified in the ASTM standards, but usually it is added to concentrations in the range of 600–1200 ppm, and this has to be agreed between producer and consumers High O concentrations (O > 2000 ppm) reduce the ductility of the alloys; therefore, O additions above 1500 ppm are not recommended In addition, O atoms interact with the dislocations at moderate temperatures, 223 leading to age-strengthening phenomena in temperature ranges depending on strain rate.20 The oxygen in solid solution in a-zirconium is an interstitial in the octahedral sites In the Zr–O system, the only available stable oxide is ZrO2 A monoclinic phase is stable at temperatures up to about 1200 C, above which it transforms to a tetragonal structure The impact on corrosion of the different phases of ZrO2, according to temperature and pressure is discussed in Chapter 5.03, Corrosion of Zirconium Alloys Tin tends to extend the a-domain, and has a maximal solubility in the hcp Zr of wt% at 940 C (Figure 5) It was originally added at concentrations of 1.2–1.7% to increase the corrosion resistance, especially by mitigating the deleterious effects of nitrogen The amount of Sn needed to compensate the effect of 300 ppm of N is about 1% of Sn However, in N-free Zr, Sn has been observed to deteriorate the corrosion resistance Therefore, the modern trend is to reduce it, but only slightly, in order to maintain good creep properties.21 Iron, chromium, and nickel, at their usual concentrations, are fully soluble in the b-phase (Figure 6) However, in the a-phase, their solubilities are very low: in the region of 120 ppm for Fe and 200 ppm for Cr at the maximum solubility temperature.22 In the pure binary systems, various phases are obtained: ZrFe2 and ZrCr2 are Laves phases with cubic or hexagonal structure, while Zr2Ni is a Zintl phase with a body-centered tetragonal C16 structure These precipitates are called the Second Phase Particles (SPPs) In the Zircaloys, the Fe substitutes for the corresponding TM and the intermetallic compounds found in Zircaloy are Zr2(Ni,Fe) and Zr(Cr,Fe)2 2000 g-ZrO2 1800 b-Zr 1500 a-Zr 1000 aЈ 500 aЈЈ® 10 a2ЈЈ 20 ¬ aЈЈ ¬ aЈЈ 30 40 50 Atomic percent oxygen Figure Zr–O binary phase diagram b-Zr 1400 b-Zr + Zr5Sn3 1200 1000 800 a-Zr Zr4Sn b-ZrO2 2000 Temperature (ЊC) 1600 a-ZrO2 Temperature ( ЊC) L L 2500 600 400 200 60 70 10 15 20 Atomic percent tin Figure Zr–Sn binary phase diagram 25 30 224 Zirconium Alloys: Properties and Characteristics 2000 1800 Temperature (ЊC) 1600 L 1400 b-Zr 1200 1000 800 ¬ FeZr3 FeZr2 600 ¬ a-Zr 400 200 (a) 10 15 20 25 Atomic percent iron 1800 ¬ g-Cr2Zr 1600 1600 b-Zr Temperature (ЊC) ¬ 1400 a-Cr2Zr ¬ Temperature (ЊC) 40 1200 1000 ¬ a-Zr 1400 L ¬ b-Zr 1200 1000 800 600 600 ¬ a-Zr 400 400 200 NiZr2 L 1800 200 (b) 35 2000 2000 800 30 10 20 30 40 50 60 70 Atomic percent chromium (c) 10 15 20 25 30 35 40 Atomic percent nickel Figure Zr-rich site of the Zr-transition metal binary phase diagram: (a) Zr–Fe, (b) Zr–Cr, (c) Zr–Ni The formation of these precipitates, and more complex ones in industrial alloys, is analyzed in detail for the control of the corrosion behavior of the Zircaloys Indeed, a strong correlation has been observed between precipitate size distributions and corrosion kinetics, the behavior being opposite for BWRs and PWRs A better uniform corrosion resistance is obtained for Zircaloys used in PWRs if they contain large precipitates, while better resistance to the localized forms of corrosion is seen in BWRs in materials that have finely distributed small precipitates.23,24 With an increase in the particle diameters from 0.05 to 0.1 mm or higher, the in-pile corrosion of Zircaloy cladding diminishes appreciably However, nodular corrosion may occur in BWR cladding with a further increase in the particle diameters above about 0.15 mm25 (Figure 7) Due to the low solubility of the transition metals (Fe, Cr and Ni) in the Zr matrix, coarsening of the precipitates, after the last b-quench, occurs at very low rates, during the intermediate annealing heat treatments, following each step of the rolling process Therefore, the precipitate growth integrates the thermal activation times of each recovery, and their temperatures and durations can be used to control the size of SPPs This integrated coarsening activation time is referred as the ‘A ’ or ‘SA ’ parameter The A-parameter calculates the integral of the activation processes for the different anneal durations and temperatures The annealing parameter is defined as A ¼ Si (ti exp (ÀQ/RTi), where ti is the time (in hours) of the ith annealing step, at temperature Ti (in K); Q/T is the activation temperature of the process involved The activation energy for the process should have been taken as the one controlling the coarsening, that is, the diffusion However, as the early studies were undertaken with the aim of improving the corrosion resistance, an unfortunate practice has been induced to take 40 000 K as the value of Q/T A more correct value would be Zirconium Alloys: Properties and Characteristics 2000 10 1600 Temperature (ЊC) BWR Relative corrosion rate L 1800 In-reactor PWR 1400 (b-Zr, b-Nb) 1200 1000 800 600 ¬ a-Zr 400 0.8 0.02 200 Out-of-pile 30 10 0.02 20 40 60 Atomic percent niobium 80 100 Figure Zr–Nb binary phase diagram 500 ЊC/16 h 225 350 ЊC/ 1a 0.1 Average diameter of precipitates (mm) 0.8 Figure Effect of precipitate size on the corrosion kinetics of Zircaloys Reproduced from Garzarolli, F.; Stehle, H Behavior of structural materials for fuel and control elements in light water cooled power reactors, IAEA STI/PUB/721; International Atomic Energy Agency: Vienna, 1987; p 387 32 000 K, which fits very well with the recrystallization kinetics The influence of the A-parameter on the corrosion of Zircaloy is discussed in more detail in Chapter 5.03, Corrosion of Zirconium Alloys High resistance to uniform corrosion in PWR is obtained for the A-parameter close to (1.5–6.0) Â 10À19 h In BWR, the A-parameter value for the Zircaloy-2 cladding in BWR has to be in the range (0.5–1.5) Â 10À18 h (Figure 7).25 This corresponds to precipitates larger than 0.18 mm The SA approach has been developed for the Zircaloys and is clearly not applicable for other alloys, such as the Zr–Nb alloys Niobium (columbium) is a b-stabilizer that can extend the bcc domain to a complete solid solution between pure Zr and pure Nb at high temperatures (Figure 8) A monotectoid transformation occurs at about 620 C and around 18.5 at.% Nb The solubility of Nb in the a-phase is maximal at the monotectic temperature, and reaches 0.65% Water b-quenching of small pieces leads to the precipitation of a0 martensite supersaturated in Nb Tempering at intermediate temperature results in b-Nb precipitation within the a0 needles and subsequent transformation of a0 into a When quenching is performed from an a ỵ b region, a uniform distribution of a- and b-grains is obtained, and the Nb-rich b-phase does not transform By aging at temperatures in the range of 500 C, the metastable Nb-rich b-phase can be decomposed into an hcp o-phase This gives a sharp increase in mechanical strength because of the fine microstructure obtained by the b-o transformation.26 In the usual form of the Zr2.5% Nb, the cold work condition after a ỵ b extrusion and air-cooling, the microstructure consists of Zr grains with layers of b-Nb rich phase (close to eutectoid composition) Owing to the affinity of Fe for the b-phase, most of this element is found in the minor b-grains These b-grains are metastable and decompose, upon aging, to a mixture of a-Zr and pure b-Nb The Nb dissolved in the a-hcp Zr phase is itself metastable and the irradiation-induced precipitation of the supersaturated Nb solid solution is believed to be the origin of the improvement in corrosion resistance under irradiation of these alloys.27 In the case of Zr–1% Nb used for VVER and RBMK, or M5® in PWRs, the concentration of Nb in the Zr matrix after processing corresponds to the maximum solubility near the monotectoid temperature, which is higher than the solubility at the service temperature Owing to the slow diffusion of Nb, the equilibrium microstructure cannot be obtained thermally However, the irradiation-enhanced diffusion allows precipitation of fine b-Nb needles in the grains after a few years in reactors.28 226 Zirconium Alloys: Properties and Characteristics Sulfur has recently been observed to be extremely efficient in improving the creep resistance, even at concentration as low as 30–50 ppm This chemical species, formerly not considered as important, is now deliberately added during processing to reduce the scatter in behavior and to improve the high temperature mechanical properties.29 The efficiency of such low concentrations on the creep properties has been explained by the segregation of the S atoms in the core of the dislocations, changing their core configurations It does not affect the corrosion properties.30 In the case of complex alloys, other thermodynamical interactions are expected and intermetallic compounds including three or four chemical elements are observed The chemistry and the crystallography of these phases may be rather complex Two examples will be given of the complex structure and behavior of these intermetallics For the Zr–Cr and Zr–Ni binary alloys, the stable forms of the second phase are Zr2Ni or ZrCr2 These phases are effectively the ones observed in the Zircaloys, with Fe substituting for the corresponding TM Therefore, the general formulae of the intermetallic compounds in Zircaloys are Zr2(Ni,Fe) and Zr(Cr,Fe)2 The crystal structure of the Zr(Cr,Fe)2 precipitates is either fcc (C15) or hcp (C14), depending on composition and heat treatment Both structures are Laves phases, with characteristic stacking faults as seen in Figure The equilibrium crystallographic structure is dependent upon the Fe/Cr ratio, cubic below 0.1 and above 0.9, and hexagonal in the middle Under irradiation, these precipitates transform to amorphous state and release their Fe in the matrix, with strong impact on corrosion behavior under irradiation.31 In the Zr–Nb–Fe ternary, other intermetallic compounds can be observed (Figure 10): the hexagonal Zr(Nb,Fe)2 phase and the cubic (Zr,Nb)4Fe2.32 Although of apparent similar composition, the two phases are indeed different: Nb can substitute Fe in the hexagonal phase, while it will substitute Zr in the cubic phase In these alloys, due to the slow diffusion of Nb, metastable phases are often present and the equilibrium microstructure after industrial heat treatments may be far from the stable one Therefore, the final microstructure is strongly dependent on the exact thermomechanical history mm 200 nm Figure Microstructure of recrystallized Zry-4: Zr(Fe,Cr)2 precipitates in the Zr(Sn–O) matrix (TEM at two different scales) 0.1 α-Zr + cub Fe (wt%) 0.08 α-Zr + cub + hex α-Zr + hex β-Zr + hex β-Nb + hex + cub Hex + cub β-Zr phase boundary Domain limit Domain limit α-Zr + Zr3Fe + cub 0.06 0.04 α-Zr + β-Nb + hex α-Zr + β-Zr (metastable) + β-Nb + hex 0.02 α-Zr + β-Nb α-Zr 0 0.2 0.4 0.6 Nb (wt%) 0.8 Figure 10 Zr-rich corner of the Zr–Nb–Fe ternary phase diagram at 580 C 1.2 Zirconium Alloys: Properties and Characteristics In addition, the low solubility of these elements at operating temperatures drastically reduces the diffusion kinetics and requires more than a year to reach equilibrium at 450 C, in the absence of irradiation.33 Other minor constituents are often found in the form of precipitates Among them are the carbide fccZrC and silicides or phosphides of various stoichiometries (Zr3Si, ZrSi2, ZrP, and Zr3P) that act as nucleation sites for the b ! a-phase transformation during quenching and, therefore contribute to control the a-platelets thickness and density 2.07.4.2 Industrial Alloys The zirconium alloys in use today for nuclear applications are limited in number: besides pure Zr, only four alloys are currently listed in the ASTM standards for Zr ingots for nuclear applications (ASTM-B350) Those are shown in Table The first three are used for cladding and structural materials, such as guide tubes and channel boxes in PWRs and BWRs and structural materials in CANDU reactors, while grade R 60904 is used exclusively in pressure tubes for CANDU reactors For cladding tubes, only Zircaloy-2 and -4 are listed in the applicable standard (ASTM B-811) Alloys of more recent use such as ZIRLO®, M5®, E110, or E635 are now of common use in light water reactor cladding, but are not considered for ASTM designation in the near future Due to the limited market of cladding tubes for nuclear reactors, and the small number of tube producers or fuel vendors, the exact chemistry, processing routes, or mechanical properties are usually agreed mutually between the contracting parties Historically, the first Zircaloy was conceived in the United States as a 2.5% Sn alloy Owing to its poor long-term corrosion behavior, the tin content was reduced roughly by a factor of A fortuitous Table Zircaloy-2 Zircaloy-4 M5 E110 Zr 2.5Nb E125 Zirlo E635 227 contamination of one melt by stainless scraps showed the drastic improvement induced by small additions of Fe, Cr, and Ni, the constituents of the austenitic stainless steels Systematic composition variations to optimize the alloy introduced the Zircaloy-2 The capture of a significant amount of hydrogen by the alloy during corrosion was attributed to the presence of nickel Its replacement by an equivalent amount of iron and chromium led to the Zircaloy-4.34 Zr–Nb alloys were developed in Canada, Russia, and the United States, with initial high Nb concentrations (up to 4%) For the claddings of BWRs, they showed poor behavior and the Zr–Nb alloys development was stopped soon in the United States Zr–2.5Nb was quite satisfactory for the pressure tubes, due to its low hydrogen pick-up during operation, and the engineering optimization of this alloy was continued in Canada and Russia It remains the reference alloy for pressure tubes, in CANDUs Zr–1% Nb has been developed for cladding in these countries and behaved very satisfactorily in VVER A renewal of interest in such alloys in the western world in the 1990s led to the development of a Zr–1% Nb alloy, with controlled additions of Fe and S The M5® alloy is now of regular use in PWRs, with excellent corrosion resistance, compared to the former Zry-4.35 ‘Quaternary’ alloys were conceived as a mixture of the Zircaloys and the Zr–Nb alloys, hoping to conserve the specificity of each of them in addition to the different alloying elements: Niobium for the resistance to hydrogenation during corrosion Tin for the corrosion resistance by reducing the dependence on deleterious impurities Iron also for corrosion resistance by mitigating the dependence on the coolant temperature The results appear to be in-line with the expectations for these alloys, which can be considered as variants Composition of the Zr alloys of industrial use (concentration in wt% or ppm) ASTM Sn R 60802 R 60902 1.2–1.5 1.2–1.7 R 60904 1.2 Nb 0.8–1.2 2.5–2.7 2.5–2.6 1 Fe Cr Ni O 0.07–0.2 0.18–0.24