Article

Effect of the addition of Cu on irradiation induced defects and hardening in Zr-Nb alloys

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Abstract

The irradiation hardening, growth, and creep of Zr alloys are strongly influenced by the presence of alloying elements. Cu is added to some Zr alloys due to its ability to form second phase particles which can cause strengthening. However, how the addition of Cu affects the microstructure and mechanical properties under irradiation remains unclear. In the current study, two Zr alloys, Zr-2.5Nb and Zr-2.5Nb-0.5Cu were selected and irradiated to 0.5 dpa and 5 dpa at 573 K with 5 MeV self-ions. The evolution of microstructure and hardness were characterized by Transmission Electron Microscopy and nanoindentation. Results show that the existence of Cu in the Zr alloys significantly altered the irradiated microstructure and resultant hardness. The addition of Cu resulted in a slight reduction of the irradiation loop size and density in both α and β phases, and its presence in β-Zr notably delayed the precipitation of ω phase. The redistribution of Fe from β phase to α phase was also slower in Zr-2.5Nb-0.5Cu than in Zr-2.5Nb. Irradiation caused hardening in both alloys, however, the hardening in Zr-2.5Nb-0.5Cu was not as significant as in Zr-2.5Nb due to the delayed ω precipitation.

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... Zr is called "the first metal of the atomic age" due to its widespread applications [10]. Therefore, there has been extensive research on the Zr alloys such as Zr-Al [11], Zr-Ti [12], Zr-Nb [13], Zr-Nb-Cu [14,15], and Zr-Nb-Mo [16] that exhibit high strength and elongation at different temperatures. It is also necessary to join the structural components of the Zr alloys to increase their stability and lower the difficulty of manufacturing. ...
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( ) Research on the diffusion bonding of ( ) similar joints of zirconium (Zr) alloys is limited as compared to that on the diffusion bonding of ( ) dissimilar joints. The similar Zr alloys are difficult to bond together owing to its high melting point; however, an added interlayer can solve this problem. This study demonstrated the successful vacuum diffusion bonding of the Zr-2.5Nb (Zr705) alloy with a Cu interlayer at 900–960 °C and analyzes the microstructure and mechanical properties of the diffusion-bonded joints. The layered morphology of the joints at 900 °C and 920 °C is attributable to the formation of intermetallic compound layers ( ). In contrast, no such formation was observed at 940 °C and 960 °C owing to the complete diffusion of Cu atoms into the Zr substrate ( ). The base material at the bonding temperatures of 900 °C and 920 °C exhibited two different microstructures i.e., a Widmanstätten microstructure near the bonding interface and a duplex microstructure away from the bonding interface. However, the ( ) temperatures at 940 °C and 960 °C exhibited an entirely Widmanstätten microstructure ( ). The tensile strength of the joints increased with the ( ) bonding temperature, from 78 MPa at 900 °C to a maximum of 603 MPa at 960 °C (joint efficiency = 104.8 %), and the elongation first increased and then decreased with the increasing temperature; at 940 °C, it reached 54% that of the original Zr705 alloy.
... range of characteristics such as the matrix composition; grain sizes, shapes and relative orientations; composition and distribution of the phases (including metastable); dislocation density; and texture. It is important to emphasize that all of these parameters critically affect the functional properties of the zirconium alloys (Zavodchikov, 2008;Song et al., 2009;Markelov, 2010;Isaenkova, 2011;Shoesmith and Zagidulin, 2011;Allen et al., 2012;Poletika, 2012;Shishov, 2012;Nikulin et al., 2012;Kim et al., 2013;Adamson et al., 2013;Lumley et al., 2013;Zavodchikov et al., 2012;Evans, 2014;Sarkar and Murty, 2015;Motta et al., 2015;Corvalan et al., 2015;Tian et al., 2015;Sudhakar Rao et al., 2015;Matsunaga et al., 2015;Kumar et al., 2015;Rajpurohit et al., 2016;Jha et al., 2016;Nguyen et al., 2019;Király et al., 2019;Dong et al., 2019). In this section, a brief overview of the structures, the properties, and the applications of the up-to-date commercial zirconium alloys (Tables 1-3) is given based on the data published in (Zaymovskiy et al., (1994) ;Reshetnikov, 1995b;Properties and applicatio, 2020;Zircadyne® 702 and 705 and h, 2020;Sparkovich, 2005;Sutherlin, 2020;Zirconium in sulfuric aci, 2020a;Zirconium in sulfuric aci, 2020b;Zirconium in hydrochloric, 2020;Rudling et al., 2008;Quality and reliability a, 2015;Nikulin, 2007;Adamov, 2005;Kalin, 2008;Nikulina and Malgin, 2008 Rudling et al., 2007). ...
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The contribution of the elastic interaction between dislocations moving in slip planes and of randomly distributed prismatic dislocation loops and of cavities to the critical shear stress in f.c.c. metals is estimated. The results are applied to quench hardening in aluminium.
Article
Transmission electron microscopy and diffraction have been used to study the isothermal omega transformation in Zr-Nb alloys. A modulated structure of omega particles in a beta matrix is developed when alloys in the composition range 12 to 25% Nb are aged at appropriate temperatures. The omega particles are cuboidal, with {100}β faces, and are aligned in 〈100〉β directions in a three-dimensional array. The average particle size is 1000 Å, and the interparticle spacing is a fraction of this value. Both the particle shape and the alignment can be interpreted in terms of the degree of precipitate-matrix lattice mismatch and elastic strain considerations. The critical factor is the elastic anisotropy of the beta matrix, having soft 〈100〉 directions. The development of the modulated beta-omega microstructure involves selective nucleation and growth of particles at preferred locations with respect to existing particles. The growth kinetics during this selective nucleation and coarsening process obey a t1 3 law. The effect of quenched omega phase formed prior to aging is important only in lean alloys.
Article
In a CANDU reactor, the fuel bundles and primary coolant are contained within Zr–2.5Nb pressure tubes that are approximately 6.3m in length, have an internal diameter of 104mm and a wall thickness of 4.2mm. The Zr–2.5Nb pressure tubes are nominally extruded at 815oC, cold-worked 27%, and stress relieved at 400oC for 24h, resulting in a structure consisting of elongated grains of hexagonal-close-packed α-Zr, partially surrounded by a thin network of filaments of body-centred-cubic β-Zr. These β-Zr filaments are metastable and contain about 20% Nb. The stress-relief treatment results in partial decomposition of the β-Zr filaments with the formation of hexagonal-close-packed ω-phase particles that are low in Nb, surrounded by a Nb-enriched β-Zr matrix. A temperature–time–transformation (TTT) diagram has been developed for the β-phase in Zr–2.5wt%Nb pressure tubes. The results show that the morphology and/or physical state of the β-phase has a significant effect on the transformation behaviour compared with a bulk Zr–20wt%Nb alloy. This means that a specific TTT-diagram is required to describe the behaviour of these engineering components.
Article
The decomposition of the β-phase in Zr-rich Zr-Nb alloys by three processes viz., ω formation, α-formation and hydride precipitation has been examined. In the Zr-20Nb alloy, ω-formation has been examined after thermal treatment as well as after electron irradiation and a comparison has been made between the kinetics of ω-phase formation under these two conditions. The morphology of the α-precipitates and their internal structures has been found to depend upon the type of thermal treatment with step quenching from the β-phase field leading to an allotriomorphic morphology and quenching and aging leading to internally twinned Widmanstätten α. The different morphologies obtained due to change in thermal treatment and composition of the ZrNb alloys has been rationalized. Hydride formation has been examined in α-Zr, β-Zr and in α + β microstructures. A comparison has been made between the mechanism of formation of hydride phase in these three types of microstructures and their morphology and internal structures have been explained.
Article
The effect of alloying additions and impurities on radiation damage in Zr-alloys has been studied using a high voltage electron microscope (HVEM). The electron damage results have been compared with neutron damage results for the same materials. There is some qualtitative agreement between the two techniques in that the relative susceptibility to c-component loop formation is the same in both the neutron and electron case. Cavity formation, however, occurs more readily during electron irradiation. Because of the relationship between c-component loop formation and accelerated growth, these results show that electron irradiation using a HVEM can provide a preliminary indication of the susceptibility of Zr-alloys to accelerated growth prior to their use in nuclear reactors.
Article
An analytical transmission electron microscopy study of two-phase (α-β) structures in a Zr-2.5 wt% Nb pressure tube alloy was used to follow the distribution of Nb and Fe as a function of alloy heat treatment and tube processing. The presence of Fe (~ 0.1 wt%) modifies the (α-β) phase equilibria as Fe is a β-stabilizing element. Significant segregation of Fe to βZr and βNb structures was measured and shown to be in good agreement with the prediction of a graphical method used to find the phase boundaries in the ternary diagram (at the Zr corner) from the binary data. Segregation of Fe at α-α subboundaries was found in the as-extruded and cold-worked plus stress-relieved treatments. The Fe profiles, measured across subboundaries, indicated a strong segregation of Fe to the core or near-core regions of the dislocations comprising the boundary structure.
Article
Recent and current work has shown that self- and substitutional diffusion in α-Zr is enhanced by the presence of the ultra-fast, interstitially-diffusing solute, Fe. The bulk of the data indicates that the diffusion coefficients scale directly with the Fe concentration. This work considers the nature of the Fe-enhancement mechanism in terms of vacancies strongly bound to either substitutional or interstitial Fe atoms. Experimental evidence from diffusion experiments, as well as from electron microscopy and positron annihilation spectroscopy investigations, is considered. It is concluded that the experimentally observed substitutional diffusion enhancements are most readily interpreted in terms of transport via an interstitial Fe/vacancy pair.
Article
The structure of Zr-2.3%Nb and Zr-5.5%Nb alloy martensites on tempering at different temperatures in the range of 350 to 600°C was studied by optical and transmission electron microscopy. The equilibrium β-niobium phase () was found to be the precipitating phase on tempering the Zr-2.3%Nb martensites at temperatures up to 500°C and the Zr-5.5% Nb martensites at temperatures upto 450°C. Precipitation of the metastable phase of the monotectoid composition (Zr20%Nb) was observed to occur on tempering the Zr-2.3%Nb alloy at 550 and 600°C and the Zr-5.5%Nb alloy at 500 and 550°C. On tempering the latter alloy at 600°C, the martensite was found to revert back to the supersaturated β phase, which subsequently decomposed into a mixture of the α and the phases. These observations have been explained on the basis of hypothetical free energy versus composition diagrams. The_orientation relation of the precipitates with respect to the α phase was found to be as follows: ; . It was also seen that a precipitate forming at a twin boundary maintains equivalent orientation relations with the two adjacent twin related portions.
Article
Le zircaloy, utilisé à la fois comme matériau de structure et comme combustible et élément poison servant de revêtement dans les corps de réacteur, refroidi à l'eau est soumis à une déformation contrôlée par l'effort qui, dans de nombreux cas peut atteindre le domaine plastique. Un appareillage a été réalisé et construit pour réaliser des essais de traction en réacteur dans une loupe à eau à 282 °C (540 °F), à des vitesses de déformation comprises entre 5 × 10−6 et 10−3 par heure. Les essais de traction du matériau dans l'état non irradié et irradié étaient réalisés à 282 °C et 315 °C pour des vitesses de déformation comprises entre 3 × 100 et 5 × 10−6 par heure. Les échantillons étaient prélevés dans le sens longitudinal et le sens transversal par rapport à la direction de laminage d'une plaque de zircaloy-4. Au total, 7 essais en réacteur ont été conduits à 282 °C: 2 échantillons longitudinaux aux vitesses de déformation de 10−5 et 5 × 10−6 par heure, et 5 échantillons transversaux aux vitesses de déformation de 1 × 10−4à 5 × 10−6 par heure. La plupart des échantillons ont été essayés jusqu'à une déformation totale comprise entre 1 et 3%. Les résultats actuels des essais en réacteurs indiquent que dans un environment neutronique et dans l'intervalle de vitesses de déformation étudiées, le zircaloy-4 a un indice apparent de sensibilité à la vitesse de déformation m = 0,23, au lieu de 0,03 pour le zircaloy-4 non irradié et après irradiation. Comme avec le zircaloy-4 non irradié, la limite élastique à 0,2% en réacteur et dans la direction transverse était plus élevée que dans la direction longitudinale, la limite élastique à 0,2% en réacteur, à la vitesse de déformation de 5 × 10−6 par heure étaient considérablement plus faibles que pour des essais comparables sur un échantillon après irradiation.
Article
The effect of nuclear radiation on the mechanical properties of copper was studied. It was found that the yield stress, which is substantially increased by the radiation, increases as the cube roet of the flux. A strong temperature dependence of the yield stress of irradiated copper was observed with the yield stress being given by a function similar to sigma = A - BT1/2 above 40 deg K. A Luders band with slip lines of very large step height was associated with the enhanced yield stress at small strains. At large strains the phenomenon of overshoot was observed. The annealing kinetics of the radiation hardness were also studied in the temperature range from 25 to 700 deg K. Little or no annealing was observed in the region below 80 deg K. In the region from 80 to 300 deg K, approximately 20% of the yield stress was recovered, with the remainder annealing in the range from 600 to 700 deg K. These results are discussed in terms of the possible mechanisms by which the hardening can occur. While by no means conclusive, these duta support a dislocation locking mechanism. On the other hand a very close analogy exists between radiation hardening and the hardening which arises from the addition of impurities, e.g., the hardening in alpha brass. The correlation between work hardening and radiation hardening appears to be quite small. (auth)
Article
The effects of some heat treatments on the hardness of Zr--2.5 wt% Nb, ; and the effects of neutron irradiation at 50 deg C and at 250 deg C on its ; tensile and impact properties are described. Post-irradiation damage recovery ; data are also presented. It is concluded that the alloy's mechanical properties ; both before and after irradiation are significantly better than those of the ; Zircaloys. The heat-treated Zr-Nb alloy appears to be metallurgically stable ; under irradiation. (auth);
Article
Over forty years of in-reactor testing and over thirty years of operating experience in power reactors have provided a broad understanding of the in-reactor deformation of cold-worked Zr–2.5Nb pressure tubes, and an extensive data-base upon which to base models for managing the life of existing reactors and for designing new ones. The effects of the major operating variables and many of the metallurgical variables are broadly understood. The deformation is often considered to comprise three components: thermal creep, irradiation growth and irradiation creep. Of the three, irradiation growth is best understood – it is thought to be driven by the diffusional anisotropy difference (DAD). It is still not clear whether the enhancement of creep by irradiation is due to climb-plus-glide (CPG), stress-induced preferred absorption (SIPA) or elasto-diffusion (ED). The least understood area is the transition between thermal creep and irradiation where the fast neutron flux may either suppress or enhance the creep rate. The three components are generally treated as additive in the models, although it is recognized that this is only a crude approximation of reality. There are still significant gaps in our knowledge besides the thermal- to irradiation-creep transition, for example, the effect of Mo which is produced from Nb by transmutation in the thermal neutron flux is not known, and on-going work is required in a number of areas. This paper reviews the current state of knowledge of the in-reactor deformation of cold-worked Zr–2.5Nb pressure tubes, and highlights areas for further research.
Article
Annealed Zircaloy-2 specimens irradiated to at ≈425 K were tensile-tested, together with unirradiated material, between 298 and 673 K in vacuum. The surface and microstructure of deformed specimens were observed using projector, optical and transmission electron microscope. Metallographie examinations or irradiated samples showed that localized deformation bands occur at intervals during deformation to the ultimate tensile stress between 473 and 623 K, while in the room temperature deformation, the inhomogeneity in metallographic features was characterized by microscopic dislocation channeling structure, without showing localized band on the surface of deformed specimen. From the shape of the stress-strain curves and metallographic features, it was concluded that the first localized band occurred after a small amount of strain, resulting in yielding and that the flow stress increased monotonically with further yielding. Evidences from temperature dependent mechanical properties of irradiated and unirradiated samples indicate that the radiation-anneal hardening phenomena are of significance in a range of temperature at which localized bands are formed, particularly pronounced at 553–593 K.
Article
Copper single crystals have been simultaneously strengthened by solved gold-atoms and by incoherent SiO2-particles. The critical resolved shear stress (CRSS) τs of a binary copper–gold solid solution and the CRSS τt of the same solid solution additionally strengthened by SiO2-particles have been measured in the temperature range 90–283 K. From the experimentally established temperature dependences of τs and τt and from the theoretically known one of τp (=CRSS if the SiO2-particles are the only strengtheners present), the function τt(τs,τp) has been derived: τt=(τks+τkp)1/k, with k≈1.8. This result is at variance with a linear relationship suggested earlier by Ebeling and Ashby [Phil. Mag. 13 (1966) 805]. The bearings of the present findings on the evaluation of experimental data on dispersion strengthening are evident.
Article
Tracer diffusion of 64Cu in α-Zr single crystals has been measured in the temperature range (615-860)°C. The temperature dependences of the 64Cu diffusion coefficients in directions parallel to and perpendicular to the c axis are given by D∥=0.40e-1.54eV/kT and D⊥=0.25e-1.60eV/kT cm2/sec, respectively. The results are discussed in terms of an interstitial diffusion mechanism.
Article
The irradiation damage configuration and yield stress response obtained in the c.p.h. α’and α phases in a Zr-2·7% Nb alloy after fast neutron irradiation have been studied as a function of pre-irradiation solute distribution. Largest yield stress increments were obtained for specimens irradiated with the solute initially in metastable solution, pre-irradiation ageing treatments reducing the yield stress response. Electron microscope studies have shown that the irradiation damage configuration was also determined by the solute super-saturation in the alloy phases, maximum strengthening being associated with a population of defect clusters or loops of < 30 Å diameter. A mechanism for the defect stabilization effect is proposed.
Article
Microhardness testing is an efficient means of assessing the mechanical properties of many materials, and is especially convenient for irradiated samples because of the small sampling volume requirement. This paper provides correlations between hardness and yield stress for both irradiated austenitic and ferritic steels by combining existing data in the open literature. For austenitic stainless steels, seven data sets were assembled and the change in yield stress was determined to simply be the change in hardness times a factor of 3.03. For the pressure vessels steels, five studies containing both hardness and yield stress data were combined. In ferritic steels, the correlation factor between change in yield stress and change in hardness was found to be 3.06. The similarity in correlation factors for austenitic and ferritic steels is consistent with previous theoretical and experimental results.
Article
The influence of irradiation (3.58 × 1024n/m2; E > 1 MeV at 583 K for 52 days) has been compared to the effects of thermal exposure alone (52 days at 583 K) on the redistribution of Nb and Fe in an as-extruded Zr-2.5 wt% Nb pressure tube alloy. The major change associated with the thermal treatment is a decomposition of the retained β-Zr phase, to form ω. As the ω-phase grows, it rejects Nb and Fe to the surrounding β-Zr. No significant changes were noted in the Nb and Fe levels segregated at α-α grain boundaries on heat treatment. Similar effects were observed in the irradiated samples, but in addition the retained β-phase is depleted of Fe. At the same time Fe is detected in the α-matrix of the irradiated specimen. Although analytical electron microscopy in a dedicated STEM instrument does not lead to reliable estimates of the absolute concentration of Fe in the irradiated sample, a series of experiments are described which show that Fe is present in the α-matrix of the alloy.
Article
The indentation load-displacement behavior of six materials tested with a Berkovich indenter has been carefully documented to establish an improved method for determining hardness and elastic modulus from indentation load-displacement data. The materials included fused silica, soda–lime glass, and single crystals of aluminum, tungsten, quartz, and sapphire. It is shown that the load–displacement curves during unloading in these materials are not linear, even in the initial stages, thereby suggesting that the flat punch approximation used so often in the analysis of unloading data is not entirely adequate. An analysis technique is presented that accounts for the curvature in the unloading data and provides a physically justifiable procedure for determining the depth which should be used in conjunction with the indenter shape function to establish the contact area at peak load. The hardnesses and elastic moduli of the six materials are computed using the analysis procedure and compared with values determined by independent means to assess the accuracy of the method. The results show that with good technique, moduli can be measured to within 5%.
Article
A study of the plastic deformation of metals by hard indenters shows that the indentation hardness is essentially a measure of the plastic yield stress of the metal. With pyramidal (and conical) indenters the hardness, because of geometric similarity, is independent of the size of the indentation. With spherical indenters this is not so, the hardness increasing with size of indentation. From the increase in hardness with load, a semi-quantitative estimate may be made of the work-hardening characteristics of the metal. In the scratching and indentation of brittle solids such as minerals it is shown that the high hydrostatic pressures developed around the deformed region are often sufficient to inhibit brittle fracture. Under these conditions the deformation is primarily plastic. For this reason there is fairly good correlation between indentation and scratch hardness since both are essentially a measure of the plastic and not the brittle properties of the solid. From this approach it is possible to provide a physical basis for Mohs' scratch-hardness scale and to show that, excluding diamond, there is a reasonable "equality of intervals" between each number on the scale.
Article
Secondary ion mass spectrometry techniques have been used to determine the terminal solid solubility (TSS) of Fe in α-Zr. Single crystals of nominally pure and Fe-doped α-Zr were annealed in the temperature range 770–1100 K to promote equilibration of Fe between surface Zr3Fe precipitates, or β-Zr(Fe), and α-Zr. The results are fair in overall agreement with a recent investigation, based on thermoelectric power measurements, but they differ in detail. In particular this work indicates two regions of temperature dependence: above 930 K the TSS (ppma) is given by CFe = 1.56 × 1010exp(−1.70 ± 0.05 eV/kT), at lower temperatures a weaker temperature dependence is associated with extrinsic effects. In addition, the eutectoid temperature is shown to lie between 1063 and 1068 K.
Article
The current understanding of defect production fundamentals in neutron-irradiated face centered cubic (FCC) and body centered cubic (BCC) metals is briefly reviewed, based primarily on transmission electron microscope observa-tions. Experimental procedures developed by Michio Kiritani and colleagues have been applied to quantify defect cluster size, density, and nature. Differences in defect accumulation behavior of irradiated BCC and FCC metals are discussed. Depending on the defect cluster obstacle strength, either the dispersed barrier hardening model or the Friedel–Kroupa–Hirsch weak barrier model can be used to describe major aspects of radiation hardening. Irradiation at low temperature can cause a change in deformation mode from dislocation cell formation at low doses to twinning or dislocation channeling at higher doses. The detailed interaction between dislocations and defect clusters helps determine the dominant deformation mode. Recent observations of the microstructure created by plastic deformation of quenched and irradiated metals are summarized, including in situ deformation results. Examples of annihilation of stacking fault tetrahedra by gliding dislocations and subsequent formation of mobile superjogs are shown. Ó 2004 Elsevier B.V. All rights reserved.
Article
Strengthening of dislocations by multiple obstacle types occurs in many engineering alloys. Theories have rationalized two different scaling laws for the total strength, tta = t1a + t2a , \tau_{t}^{\alpha } = \tau_{1}^{\alpha } + \tau_{2}^{\alpha } , with α=1 or 2, where τ 1 and τ 2 are the strengths of the two individual obstacle types. Simulations have clearly demonstrated α=2, while “friction” strengthening must correspond to α=1. Here, line-tension simulations of dislocation glide through two types of point obstacles are performed to examine the friction limit. One obstacle type is weak (critical angle approaching 180deg) but with very high density, approximately corresponding to solute strengthening; and the second obstacle type is stronger (smaller critical angle) but with lower density, approximately corresponding to forest or precipitate strengthening. Additive strengthening α=1 is obtained when the densities of the two obstacle types differ by more than a factor of ~67, while a transition to α=2 occurs with increasing density of the second obstacle. These simulations confirm the long-held metallurgical wisdom regarding additivity of solute or friction strengthening with other strengthening mechanisms and also demonstrate that apparent intermediate scaling laws with 1<α<2 can arise for a range of relative obstacle densities. Investigation of several literature experimental studies shows some agreement with the model here but quantitative comparisons remain difficult.