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Our elastic viscoplastic self-consistent (EVPSC) model is improved by including thermal strain to allow study of the behavior of a Zircaloy-2 slab under moderately large strains. The approach of introducing thermal strain effect is similar to the one developed by Turner et al. (1995) and used by Xu et al. (2008a). Various self-consistent schemes (SCSs) of the EVPSC model are then evaluated in terms of the deformation behavior of the material under different uniaxial strain paths. The material parameters for the various models are determined by fitting experimental data from uniaxial tension and compression tests along the normal direction (ND) and from uniaxial tension tests along the rolling direction (RD). The quality of the various SCS predictions is assessed based on comparisons of macroscopic deformation behavior (stress–strain curves and R-values) and microscopic mechanical response (the evolution of lattice strains and texture coefficients) between the predicted and experimental data. It is revealed that the Affine and Meff = 0.1 self-consistent models give much better performance than the Secant and Tangent models.

The pre-transition oxides formed on five different types of Zircaloy-2 alloys have been characterized in-situ using electrochemical impedance spectroscopy (EIS) in high-temperature water simulating BWR conditions at 300°C. The contribution of the oxide film properties to the impedance spectra has been distinguished from the contribution of the corrosion reaction by performing additional EIS measurements in a gas atmosphere. To obtain a correlation between the oxide film thickness, the oxide growth rate, and the impedance parameters, the EIS data have been fitted to the transfer functions derived from the Mixed Conduction Model for oxide films. As a result, the oxide growth rate of each alloy has been calculated assuming that the growth rate is proportional to the flux of oxygen vacancies through the film. The values of the oxide film thickness based on the model calculations have proved to be in good agreement with the values measured from the cross-section micrographs of the specimens using scanning electron microscopy (SEM). The measured corrosion properties have been qualitatively correlated to the secondary phase particle distributions of the five fuel cladding materials. However, to establish a quantitative correlation between the corrosion rates and the second phase particle distributions of these alloys, more data on the microstructure of the metal-oxide interface are needed.

A simple model of hydrogen diffusion and hydride precipitation was developed to estimate the radial re-distribution of hydrogen in zirconium-lined Zircaloy-2 claddings for light water reactors. Numerical calculation based on the model agreed well with experimental data and clearly showed that a slight difference of the terminal solid solubility of hydrogen in the two materials caused a significant flow of hydrogen from Zircaloy-2 to zirconium during slow cooling. The model would also evaluate the hydrogen distribution in the claddings of which the terminal solid solubility is not uniform.

The effect of irradiation temperature and alloying elements on defect clustering behaviour directly from the cascade collapse in Zircaloy-2 is examined. The in-situ ioWn irradiation technique was employed to study the formation of -type dislocation loops by Kr ion irradiation at 573 K and 773 K, while the dependence of dislocation loop formationon the presence of alloying elements was investigated by comparing with the defect microstructures of pure Zr irradiated under similar conditions. The experimentally observed temperature dependence of defect clustering was further investigated using molecular dynamics (MD) simulations near the experimental irradiation temperatures. We particularly concentrate on yield and morphology of small defect clusters formed directly from cascade collapse at very low ion doses. Smaller loop size and higher defect yield (DY) in Zircaloy-2 as compared to pure Zr suggests that the presence of the major alloying element Sn increases the number of nucleation sites for the defect clusters but suppresses the point defect recombination. MD simulations at 600 and 800 K revealed that the production of both vacancy and interstitial clusters drops significantly with an increase of irradiation temperature, which is reflected in experimentally collected DY data.

The thermodynamics of H/{\alpha}-Zr solid solution and zirconium hydride phases were studied using density functional theory. Disorder in {\zeta}, {\gamma} and {\delta} hydrides and solid solutions were modelled using a statistically significant number of randomly generated structures in combination with special quasi-random structures and solid solutions with a range of concentrations. This is used in conjunction with a calculation of thermodynamic parameters of the system, including the temperature dependent sensible enthalpy, configurational entropy and vibrational entropy of the different crystals in the system, developed from phonon density of states. It was found that precipitation of hydrides is not thermodynamically favourable for Zr-H solid solutions containing less than 300 ppm H, suggesting that a mechanism must cause local concentration of H atoms to a greater amount than found globally in experimental samples containing hydrides. Temperature drives the reaction in the direction of solution, primarily due to entropic effects. Generally, {\gamma} hydride is the most stable phase, although it is very close in energy to the {\delta}-phase. The sensible enthalpy of precipitation assists in stabilising HCP hydrides, and the configurational entropy change during precipitation favours FCC hydrides. None of the thermodynamic contributions are found to be negligible in driving precipitation., Comment: 19 pages, 11 figures, submitted to Phys. Rev. B on 22-Jan-2013

Asymmetric stress cycling leads to accumulation of plastic strain which is called as ratcheting strain. The problem is generally associated with nuclear fuel cladding materials used in nuclear power plants and in pressurized pipe lines. In the present investigation asymmetric stress controlled fatigue tests were conducted with three different parameters namely, mean stress, stress amplitude and stress rate (keeping two parameter constant and varying third parameter) to see the plastic strain accumulation and its effect on fatigue life of Zircaloy-2 at room temperature. The tests were conducted with mean stress varying from 80 to 150 MPa, stress amplitude varying from 270 to 340 MPa and stress rate varying from 30 to 750 MPa/s respectively. The experimental outcomes show that with increase in mean stress and stress amplitude, the ratcheting strain accumulation increases with reduction in fatigue life. However, increase in stress rate leads to improvement in fatigue life of the material due to small ratcheting strain accumulation. Cyclic softening is observed during initial cycles for lower values of mean stress and stress amplitude and cyclic hardening is predominant for higher values of mean stress and stress amplitude. However, cyclic hardening is observed with increase in stress rate.

Zirconium alloys used in the nuclear industry are exposed to extreme conditions undergoing high levels of irradiation damage and corrosion. Zircaloy-2 is used as nuclear fuel cladding in boiling water reactors and for the encapsulation of the spallation target in the neutron source SINQ at the Paul Scherrer Institute. As a result of oxidation, hydrogen is produced and partially taken up into the zirconium matrix. As the hydrogen reaches the solubility limit, it precipitates in the form of a zirconium hydride. A number of deleterious effects of hydrides have been extensively studied, including the embrittlement of the material and a unique cracking phenomenon, namely delayed hydride cracking (DHC). DHC occurs as hydrogen in solid solution diffuses towards locations of higher tensile stress, where it subsequently precipitates as the local solvus limit is reached. Once the hydride grows to a critical size, it fractures, which can continue to propagate in repeated steps by the same mechanism. Several studies have been performed in order to understand properties of DHC based on parameters like, alloy type, cracking temperature, hydrogen content, and cracking direction. The most important mechanical properties include the fracture toughness and cracking velocity, which governs the material failure conditions. Crystallographic properties have been investigated in order to understand how complex hydride precipitation develops within the context of DHC conditions. In this project, radial DHC is explored in irradiated and unirradiated thin-walled Zircaloy-2 nuclear fuel cladding with and without an inner liner, in addition to spallation target rod material. A unique three-point bending test, applicable also for irradiated samples in the hotlab, was developed to induce a consistent radially propagating outside-in crack. Results have provided insight into the effects of the inner liner on cracking velocity, showing correlations with the hydrogen depletion. The combined DHC and creep mechanisms, which cannot be decoupled, is coined as ‘creep-delayed hydride cracking’. As the initiating conditions for DHC are of great technical importance, the threshold stress intensity factor has been investigated, under ideal DHC conditions, where a minimum value occurs around 6 MPa√m. Quantitative analysis of the hydrogen distribution around the tip of an arrested DHC crack tip has been performed with neutron radiography at the SINQ. Trends show that the hydrogen concentration around the crack tip increases with temperature resulting likely from the larger hydrogen diffusion, and that the liner reduces the hydrogen available for DHC. Crystallographic analysis has provided insight on the hydride phases responsible for DHC with synchrotron micro X-ray diffraction, performed at the Swiss Light Source at PSI. The results directly suggest that the γ-hydride is a stable phase precipitated at low temperatures. Testing of irradiated material provided insight into the crack initiation of DHC as a result of the strongly irregular outer surface generated from the oxidation. The propagation pattern is highly irregular compared to the unirradiated material due to the embrittled nature of the cladding from irradiation damage.

Some inaccuracies are possible during the near-surface residual stress measurement of textured material by X-ray diffraction using the standard multiexposure technique. Diffraction intensities depend on factors like inclination of the sample, the rotation of the beam, and the position of detectors. Low intensity due to texture can result in increased percentage error observed stress values. Anisotropy of elastic constants in textured materials can further increase the errors in measured stress. In the present study, the effect of texture on residual stress has been studied for seamless Ti-3wt%Al-2Vwt% tubes, in pilgered and annealed conditions, by superimposing the residual stress measurements on the pole figure obtained by texture measurement. A general approach has been developed for accurate determination of the stress tensor in textured material. This consists of the selection of inclinations giving rise to high intensity peak, the selection of rotation angles symmetric to the pole figure to get similar intensity from both the detectors (where this is not possible one detector with low intensity is shut off), and the calculation of the bulk X-ray elastic constants using the single crystal elastic data and texture. Cold worked Ti-3wt%Al-2Vwt% tubes had the highest stress along the rolling direction and that is compressive in nature. The principal residual stresses in the pilgered tubes were along the rolling and transverse directions, and annealing leads to reduction and reorientation of the stresses.