Dislocations generated by zirconium hydride precipitates in zirconium and some of its alloys

The character of dislocations emitted during the precipitation of γ-zirconium hydride in zirconium, Zircaloy-2, Zr-1% Al and Zr-1% Cr has been determined. Contrast experiments in the transmission electron microscope have shown that the dislocations possess Burgers vectors ($\text{b}$) of the type $\text{1}\text{3}\text{a <11}\text{2}\text{?}\text{0>}$. Of the three possible $\text{b}$'s only one or two are observed associated with a given hydride needle, those giving a large component of $\text{b}$ along a direction perpendicular to the needle. Dislocation generation is thought to result from the dilatational misfit associated with the hydride needles. The creation of dislocations with $\text{b}$'s along the $\text{c-}\text{axis}$ is more difficult and the misfit strain along this direction is not relieved by plastic deformation.     

Diffusion in uranium alloys with zirconium and titanium

Moscow Engineering-Physics Institute. Translated from Atomnaya Énergiya, Vol. 72, No. 1, pp. 30–33, January, 1992.     

The corrosion of some zirconium alloys under radiation in moist carbon dioxide-air mixtures

In-reactor experiments are reported in which Zircaloy-2 and $\text{Zr-2}\text{1}\text{2}$ wt % Nb alloys were exposed to moist carbon dioxide-air mixtures at 300°C for periods up to ≈ 800 days. For Zircaloy-2 the corrosion was significantly enhanced by the reactor radiation but the percentage hydrogen pick-up was reduced. For $\text{Zr-2}\text{1}\text{2}$ wt % Nb alloy the effect of radiation on the corrosion rate ranged from a slight suppression to a significant enhancement, depending on the metallurgical condition of the alloy, but there appeared to be no effect on the percentage hydrogen pick-up. The effect of improving the purity of the corrodent was also studied in the absence of radiation. The results are used to predict that the corrosion and hydrogen pick-up of typical reactor pressure tubes exposed to moist carbon dioxide-air mixtures are unlikely to cause a significant deterioration of their mechanical properties.     

High temperature oxidation of some zirconium alloys in flowing oxygen

The oxidation behavior of three zirconium alloys, Zr-2.2wt%Hf, Zr-2.5wt%Nb and Zr-3wt%Nb-1wt%Sn, has been studied in flowing oxygen in the temperature range 873–1173 K to 120 ks (2000 min). Zr-2.5Nb and Zr-3Nb-1Sn showed a transition to rapid linear kinetics after initial parabolic oxidation at all temperatures. Zr-2.2Hf, on the other hand, showed this transition at temperatures in the range 973–1173 K; no transition was observed at 873 K within the oxidation times reported. Zr-2.2Hf showed the smallest weight gains, followed by Zr-2.5Nb and Zr-3Nb-1Sn. Increased oxidation rates and shorter time-to-rate transition of Zr-2.5Nb and Zr-3Nb-1Sn as compared with Zr-2.2Hf are attributed to the presence of the alloying elements Nb, Sn and Hf. Based on the Nomura-Akutsu model, Hf should delay the rate transition, while Nb and Sn lead to shorter transition times. The scale on Zr-2.2Hf was identified as monoclinic zirconia, while the tetragonal phase, 6ZrO₂ · Nb₂O₅, was contained in the monoclinic zirconia scales on both other alloys.     

The elastic anisotropy of polycrystalline aggregates of zirconium and its alloys

Hexagonal close-packed zirconium and its alloys are used extensively as components in nuclear reactors for fuel cladding, pressure tubes, etc. As a result of thermomechanical processing such components display a texture or crystallographic alignment and therefore exhibit anistotropic mechanical properties. In this paper the crystallite orientation distribution function is used to derive expressions for the elastic compliance tensor of a partially aligned aggregate of hexagonal crystallites. The angular dependence of the Young's modulus in the rolling plane is shown to depend on only three coefficients W₂₂₀, W₄₂₀ and W₄₄₀ in a series expansion of the crystallite orientation distribution function in generalised spherical harmonics. These coefficients can be obtained either by neutron diffraction or by ultrasonic techniques.     

A Comparison of the polarisation behaviour of zirconium and its alloys

Current-voltage polarisation curves have been measured for freshly abraded electrodes of Zr, Zr-2, Zr-4 in aqueous solutions, 0.1M, of H₂SO₄ and NaOH. The measurements were made at a controlled potential, changing incrementally at rates between 0.015 and 1.5 volts/min. This was done at temperatures of 20, 55 and 85°C, in a hydrogen atmosphere. The data have been analysed on the basis of a process of surface coverage followed by thickening of an anodic oxide film. Parameters to characterise both processes have been derived for each alloy. Values of the parameters for the initial surface coverage show no systematic differences between the metals tested, whereas values for ion migration in the thick film show marked differences. The temperature coefficient for a process which is attributed to the interfacial barrier is high and similar for all three metals tested, whereas the temperature coefficient for ion migration appears to be low and varies for the different metals and conditions. The charge involved in the surface rate controlling step appears to be unity and this is attributed to the hydrogen ion or the hydroxyl ion.     

Fretting-wear of zirconium alloys

Fretting tests of Zircaloy fuel sheath bearing pads in contact with zirconium alloy (Zr–2.5Nb) pressure tube specimens were conducted at temperatures varying from 25 to 315 °C. The effects of motion type and amplitude, water chemistry, fuel sheath manufacturer and pressure tube surface finish were also investigated. The effect of temperature is the most significant. The pressure tube wear coefficient in the 225–286 °C for all four motions studied is considerably greater than that above 300 °C. The fretting rate for small amplitude motion representative of flow turbulence excitation is about equal at temperatures below 150 °C and above 300 °C, but is five to ten times greater in the 250–286 °C range.     

The relation between stress-relaxation and creep for some zirconium alloys during neutron irradiation

In-reactor stress-relaxation tests on beam specimens of several zirconium alloys have been performed at 566 K in a fast neutron flux (E>1 MeV) of 2 × 10¹⁷ n/m² · s. The stress-relaxation behaviour is characterized by an initial rapid decrease in the unrelaxed stress ratio followed by a slower steady-state decrease which can be expressed by the relation,

$\text{ln}\text{(σ}\text{σ}{}_0\text{) = ? (AE)t +}\text{ln}\text{D}$

, where $\text{σ}\text{σ}{}_0$ is the unrelaxed stress at a given time t as a fraction of the initial stress σ₀, A is a flux-, material-,temperature-dependent constant, E is Young's modulus, and D is given by $\text{σ}{}_{\mathrm{p}}\text{σ}{}_0$ where σ_p approximates the initial stress decrease.The linear dependence of In $\text{σ}\text{σ}{}_0$ on time and the independence of the stress-relaxation behaviour on the initial stress implies that the creep rate in the steady-state period can be given by the expression, $\text{ε}\text{?}\text{=}\text{A}\text{σ}$. Good agreement was obtained between creep rates derived from stress-relaxation tests on experimental pressure tube materials and creep rates derived from diameter measurements of pressure tubes for identical temperature, flux and stress conditions.     

High-temperature creep and superplasticity in zirconium alloys

High-temperature (≈ 900?1400 K) steady-state creep test data on as-received zirconium alloys, Zr-1wt%Nb and Zircaloy-4 used as fuel cladding materials in light water reactors are evaluated by employing two sets of models. In particular, the focus of the paper is on the former alloy and in the two-phase coexistence region, i.e. the (α+β)-domain of the alloy. In one modeling approach, the constitutive relations for the two single phase regions (α and β) are combined through a phase transition kinetic model and a phase mixing rule; in another, a superplasticity model is used directly to calculate the creep deformation rate as a function of stress and temperature in the (α+β)-domain. The results show that the former approach is inadequate in retrodicting the experimental data, while the latter one gives a fair overall agreement. The paper describes the details of the models, the data, and derivations of the constitutive laws.     

Substructure strengthening in zirconium and zirconium-tin alloys

The substructures developed during the hot compression of α-Zr and a series of Zr-Sn solid solution alloys were examined by means of transmission electron microscopy. The materials were hot compressed in a modified Instron testing machine at constant true strain rates of 10⁻⁴ to 3 × 10⁻¹sec⁻¹. The range of testing temperatures was 625 to 825 °C. Over the subgrain size range 0.3 to 7 μm, the high temperature flow stress and subgrain size were inversely related. The increase in flow stress due to substructure formation also obeyed this type of relation. The results are consistent with a glide model of deformation in which the build-up of internal stress is due to the substructure developed by the deformation.     

Elastic constants of zirconium alloys

The dynamic elastic moduli of Zircaloy-2, Zr-1.15 wt% Ci-0.1 wt% Fe and Zr-2.5 wt% Nb have been determined over the temperature range 293–773 K. Young's modulus and shear modulus decreased linearly with increasing temperature. Poisson's ratio decreased with increasing temperature for Zircaloy-2 and Zr-2.5 wt% Nb but increased for Zr-1.15 wt% Cr-0.1 wt% Fe. The results have been compared with previous values determined both by static and dynamic techniques and with polycrystalline constants computed from single crystal elastic constants. The difference in behaviour between the three alloys is not due to differences in alloy composition but to texture effects. A relationship is derived between texture and elastic constants.     

Modelling precipitation in zirconium niobium alloys

A model has been developed to predict precipitation of β-Nb in zirconium-niobium alloys. The model considers two transformation mechanisms; in situ transformation of any retained β-Zr and homogeneous nucleation of β-Nb. The two mechanisms are allowed to operate concurrently and compete for the available solute. The model has been calibrated and tested using data in the literature and is able to reasonably reproduce these results without introducing non-physical fitting parameters. It has then been applied to predict the effects of prior β-Zr fraction, oxygen content, and temperature on the precipitation kinetics of β-Nb. These calculations predict that prior β-Zr fraction has a strong effect on the kinetics of subsequent β-Nb evolution and that oxygen content is also critical, with higher oxygen levels predicted to result in faster kinetics and shift in the peak transformation rate to higher temperatures.     

Creep behavior of niobium-modified zirconium alloys

Zirconium (Zr) alloys remain as the main cladding materials in most water reactors. Historically, a series of Zircaloys were developed, and two versions, Zircaloy-2 and -4, are still employed in many reactors. The recent trend is to use the Nb-modified zirconium alloys as the Nb addition improves cladding performance in various ways, most significant being superior long term corrosion resistance. Hence, new alloys with Nb additions have recently been developed, such as Zirlo^™2 and M5^™3. Although it is known that creep properties improve, there have been very few data available to precisely evaluate the creep characteristics of new commercial alloys. However, the creep behavior of many Nb-modified zirconium alloys has been studied in several occasions. In this study, we have collected the creep data of these Nb-modified alloys from the open literature as well as our own study over a wide range of stresses and temperatures. The data have been compared with those of conventional Zr and Zircaloys to determine the exact role Nb plays. It has been argued that Nb-modified zirconium alloys would behave as Class-A alloys (stress exponent of 3) with the Nb atoms forming solute atmospheres around dislocations and thus, impeding dislocation glide under suitable conditions. On the other hand, zirconium and Zircaloys behave as Class-M alloys with a stress exponent of ?4, attesting to the dislocation climb-controlled deformation mode.     

Localised electron transport in corroding zirconium alloys

Zircaloy-2, in the as-received and heat-treated conditions, and the binaries of Zr with Sn, Fe, Cr and Ni, were oxidised in a fused salt medium at 300 and 400° C. The electrochemical polarisation behaviour was studied at various oxide film thicknesses. Metal electrodes were evaporated on to the oxide films and the dc conduction of the alloy-oxide-metal diodes was investigated. Analysis of the data shows that the presence of the Zr-Fe intermetallic phase is associated with an enhanced localised electron transport which leads to low negative corrosion potentials and fast oxidation rates; the high temperature oxidation behaviour of zirconium alloys falls into two distinct groups depending on the presence or absence of this intermetallic phase. In proposing a model for the oxidation of zirconium alloys, information regarding the processes controlling ion and electron transport is essential. Because of the possibility of electron transport occurring through the bulk oxide and at localised second-phase precipitates by different mechanisms, it is unlikely that there is a simple relation between the electron current and potential drop across the oxide.     

Quantitative characterization of crystallographic textures in zirconium alloys

The techniques for determining inverse pole figures and direct pole figures of zirconium alloys by X-ray diffraction are summarized, and their advantages, disadvantages, and limitations are discussed. A critical review is made of the various parameters that have been used to quantify the texture in zirconium alloys. A new series of four quantitative texture numbers F_T, (SD)_T, F_A, and S, which are obtained from the direct pole figure, are proposed. Pole figures are determined for Zircaloy-2 tubing produced by three tubing manufacturers. The four texture numbers are calculated and are used to compare the textures of the three manufacturers and the through wall texture gradient of one manufacturer.     

Radiation induced creep and growth of zirconium alloys

The strain under irradiation of zirconium and its alloys is calculated within a simple rate theory approach. Network dislocations and interstitial dislocation loops with their Burgers vector oriented parallel to the crystal basal plane are assumed to climb by preferentially attracting interstitials with respect to vacancies, while the grain boundaries act as neutral sinks and absorb therefore more vacancies than interstitials. This same theory has been applied by Fainstein-Pedraza, Savino and Pedraza for modelling the irradiation growth of cold worked zirconium alloys. It is now extended by including the effect of vacancy traps and the stress induced preferential bias for interstitials of those dislocations favourably oriented with respect to an external or internal stress field. In addition, a model which allows to correlate the deformation of the individual grains with the strain of the polycrystalline specimen where they pertain is developed. The stresses induced within the same grain while it deforms inside the textured crystal are also numerically calculated. Those stresses modify the grain strain via the SIPA mechanism and the stresses-strains are then coupled. The calculated crystal deformation is strongly dependent on texture. For tubes with the c axis oriented preferentially on an axial plane, a rapid increase of the longitudinal strain rate is predicted at high doses.