Explore: Intermetallics

Oxidation kinetics at 300°C in dry oxygen of 0.5 wt% binary alloys of iron, nickel, and chromium in zirconium were determined for several surface preparations. Further, chemical profiles of the oxides as they existed on the matrix and on the precipitates were obtained by sputtering and Auger electron analysis. The appearance of "breakaway" oxidation was controlled by the surface finish of the alloy, a variable that could be used to eliminate the phenomenon for all alloys except the Zr/Ni binary, which required ?-quenching to accomplish the same purpose.

Various experimental techniques using mechanical springs for high strain sensitivity creep testing are described. A theory of stress and strain using these techniques is briefly summarized. Problems and advantages of the helicoid spring specimen technique are also analyzed. The helicoid spring specimen technique provides high efficiency for low-stress creep experiments in the self-loaded mode, while the external load mode allows extraordinary strain sensitivity. A new machine for conducting high-sensitivity torsion creep tests is presented together with some examples of creep curves obtained by the technique. The machine was designed to enable high sensitivity creep tests of inherently brittle materials such as intermetallics and ceramics.

An earlier paper (Trowse et al, ASTM STP 633) described the nodular corrosion of Zircaloy and other zirconium alloys in a steam generating heavy water reactor (SGHWR) and related environments. The observations have now been extended to more highly irradiated elements up to the design burn-up of 20 MWD/kgU mean. Maximum observed local oxide thicknesses were less than the earlier predictions of 200 ?m for SGHWR cladding at the design burn-up and are acceptable from fuel performance considerations.

In steam generating heavy water reactors (SGHWR), nodular or "patch-type" corrosion is the dominant form of attack on Zircaloy, as it is in other boiling or oxygenated reactor environments. The most severe oxidation occurs close to the stainless steel spacer grids, although nodules are seen elsewhere on cladding and on non-fueled components, a maximum thickness of 160 ?m at 20 MWd/kg uranium local burn-up having been recorded to date. The extensive data from post-irradiation examination of SGHWR fuel elements are reviewed and discussed from the points of view of determination of rate laws for oxide growth and the establishment of corrosion mechanisms.

A titanium-21 weight percent germanium eutectic alloy was unidirectionally solidified at two different rates to produce a composite material strengthened by intermetallic Ti5Ge3 fibers. The higher solidification rate produced a composite with fibers of finer diameters. The fracture behavior of this material was determined from work of fracture and impact experiments. A brittle-to-ductile transition was found, through the work of fracture experiments, to be 1150°F (621 °C). This transition was raised by some 450°F (232°C) under impact loading. The large increase in the work of fracture corresponding to this transition has been attributed to the increased ductility of the titanium matrix. It was also found in this experiment that the composite with larger fiber diameters was tougher than that with smaller fiber diameters. Macroscopic and microscopic observation of the fracture surface revealed that failure occurred in a brittle fashion below 800°F (427°C) and by normal rupture between 800 and 1200°F (427 and 649°C).

The precipitation of intermetallic compounds in Zircaloy-4 has been studied by transmission electron microscopy. The classical precipitates with a ZrCr2 lattice have been observed. In addition to these precipitates, a hexagonal phase with lattice parameters a = 7.9 Å and c = 3.3 Å can be developed by quenching from the temperature range 840 to 870°C. This phase is metastable and transforms into a zirconiumiron compound by aging at 600 to 650°C.

The production of solute enriched (or depleted) regions has been observed in Fe3Al, Cu5Zn8, and Ni5Zn21 subject to neutron irradiation. The formation of those regions is due to the separation of the alloy components in the path of the atomic collision cascade, at the thermal stage of its development. A mechanism of formation of a region of different concentration in the collision cascade is proposed. Analyses of the kinetics of recovery of the in-cascade induced defect distribution show that the annealing of that damage can lead to the redistribution of the alloy components. Enrichment of one atomic species results in the depleted zone. The preferential flux of an atomic species toward the depleted zone is determined by the difference in the atomic mobilities and the characteristics of the interstitial configurations and migration mechanisms. A computer simulation of the separation confirmed the possibility of precipitation of a new phase at a sink, such as a depleted zone, which exists for a very short period of time.

The plain and fretting fatigue behavior was studied of a ?-base TiAl intermetallics at 675°C in air. he TiAl material was produced by the XDTM process and had a nominal composition of Ti-47Al-2Nb-2Mn+0.8vol%TiB2. The contact material was Inconel 718. The reduction of fatigue strength by fretting was very small (around 20%) compared to metallic materials (3060%). The plain and fretting fatigue S-N curves were flat, which resembled the behavior of ceramic materials. The flat S-N curves are corresponding to the very steep crack growth curve reported for TiAl-based materials containing TiB2. A fretting fatigue crack nucleated in the very late stage of fatigue life (later than 0.75Nf), while it is known to nucleate in the very early stage of fatigue life (less than few percent of Nf) for steels. The fatigue lifetimes of this material were suggested to be determined by the cycles needed for nucleation of a crack rather than by crack propagation.