Effective structural unit analysis in hexagonal close-packed alloys – reconstruction of parent β microstructures and crystal orientation post-processing analysis

Materials with an allotropic phase transformation can form microstructures where grains have orientation relationships determined by the transformation history. These microstructures influence the final material properties. In zirconium alloys, there is a solid-state body-centred cubic (b.c.c.) to hexagonal close-packed (h.c.p.) phase transformation, where the crystal orientations of the h.c.p. phase can be related to the parent b.c.c. structure via the Burgers orientation relationship (BOR). In the present work, a reconstruction code, developed for steels and which uses a Markov chain clustering algorithm to analyse electron backscatter diffraction maps, is adapted and applied to the h.c.p./b.c.c. BOR. This algorithm is released as open-source code (via github, as ParentBOR). The algorithm enables new post-processing of the original and reconstructed data sets to analyse the variants of the h.c.p. α phase that are present and understand shared crystal planes and shared lattice directions within each parent β grain; it is anticipated that this will assist in understanding the transformation-related deformation properties of the final microstructure. Finally, the ParentBOR code is compared with recently released reconstruction codes implemented in MTEX to reveal differences and similarities in how the microstructure is described.

Autowave Criteria of Fracture and Plastic Strain Localization of Zirconium Alloys

Plastic deformation and fracture of Zr–1% Nb alloys exposed to quasi-static tensile testing have been studied via a joint analysis of stress-strain curves, ultrasound velocity and double-exposure speckle photographs. The possibilities of ductility evaluation through the εxx strain distribution in thin-walled parts of zirconium alloys are shown in this paper. The stress-strain state of zirconium alloys in a cold rolling site is investigated considering the development of localized deformation bands and changes in ultrasound velocity. It is established that the transition from the upsetting to the reduction region is accompanied by the significant exhaustion of the plasticity margin of the material; therefore, the latter is more prone to fracture in this zone exactly. It is shown that traditional methods estimating the plasticity margin from the mechanical properties cannot reveal this region, requiring a comprehensive study of macroscopically localized plastic strain in combination with acoustic measurements. In particular, the multi-pass cold rolling of Zr alloys includes various localized deformation processes that can result in the formation of localized plasticity autowaves. Recommendations for strain distribution division over the deformation zone length in the alloy in the pilger roll grooves are provided as well.

Phase transformation of the corrosion layers on nuclear fuel cladding. Raman spectroscopy study.

The goal of this work was the phase analysis of corrosion layers on zirconium alloys. In the environment of nuclear reactors, zirconium alloys are covered with a protective layer of zirconium oxide, which occurs in two crystalline modifications - monoclinic and tetragonal. The distribution of these phases in the corrosion layer can affect the overall corrosion rate. Raman spectroscopy was used to determine the composition of the corrosion layers. The use of this method is advantageous because the monoclinic and tetragonal phases can be easily distinguished in the spectra of the corrosion layers. In total, samples of two alloys were measured. The samples were pre-exposed at 360 °C in Li+ containing water (70 mg/l Li as LiOH) . Exposure times were between 21 d and 231 d, so the series contained both pre- and post- transition samples. The relative proportion of the tetragonal phase decreases significantly after the transient. It has also been found that the corrosion layers are highly heterogeneous in terms of the distribution of crystalline modifications.

Исследование поверхностного окисления циркония активным тепловым методом

The possibility of testing the surface oxidation of zirconium has been investigated by the method of one-way active thermal non-destructive testing based on the analysis of radiation temperatures. The emissivity of the oxidized surface of zirconium samples obtained at different annealing temperatures was estimated at various stages and heating temperatures in the infrared wavelength range. It is shown that there is a principal possibility to remotely test the oxidation process of zirconium alloys used in the nuclear industry for the manufacture of fuel elements operating in the core of nuclear reactors.

Irradiation Hardening and Microstructure Characterization of Zr -1% Nb During Low Dose Neutron Irradiation

Due to low neutron absorption cross section, high mechanical strength, high thermal conductivity and good corrosion resistance in water and steam, Zirconium alloys are widely used as fuel cladding material in nuclear reactors. During life-time of a reactor the microstructure of this alloy is affected due to, among other factors, radiation damage and hydrogen damage. In this work mechanical properties changes on neutron irradiated Zr-1wt.% Nb at low temperatures (< 100 °C) and low dose (3.5 ´ 1023 n m-2 (E > 1 MeV)) were correlated with hydrides and crystal defects evolution during irradiation. To achieve this propose, tensile tests of: 1) Non-hydrided and non-irradiated material, 2) Hydrided and non-irradiated material and 3) Hydrided and irradiated material were performed at 25 ºC and 300 ºC. Different phases, hydrides and second phase precipitates were characterized by transmission electron microscopy (TEM) techniques. For the hydrided and irradiated material, the ductility decreased sharply with respect to the hydrided and non-irradiated material, among other factors, due to the change in the microstructure produced mainly by neutron irradiation. Even if the presence of the hydride ζ (zeta) was observed, both in the irradiated and non-irradiated material, tensile tests showed that ζ-hydrides did not affect ductility, since hydrided samples are more ductile than non-hydrided samples.