Zirconium | EPFL Graph Search

Zirconium is a chemical element with the symbol Zr and atomic number 40. The name zirconium is derived from the name of the mineral zircon, the most important source of zirconium. The word is related to Persian zargun (zircon; zar-gun, "gold-like" or "as gold"). It is a lustrous, grey-white, strong transition metal that closely resembles hafnium and, to a lesser extent, titanium. Zirconium is mainly used as a refractory and opacifier, although small amounts are used as an alloying agent for its strong resistance to corrosion. Zirconium forms a variety of inorganic and organometallic compounds such as zirconium dioxide and zirconocene dichloride, respectively. Five isotopes occur naturally, four of which are stable. Zirconium compounds have no known biological role. Characteristics Zirconium is a lustrous, greyish-white, soft, ductile, malleable metal that is solid at room temperature, though it is hard and brittle at lesser purities.

Delayed Hydride Cracking in Irradiated and Unirradiated Zircaloy-2 Cladding

Aaron William Colldeweih

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 âcreep-delayed hydride crackingâ. 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 MPaâm. 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 Î³-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.

EPFL2022

Processability, microstructure and precipitation of a Zr-modified 2618 aluminium alloy fabricated by laser powder bed fusion

Christian Leinenbach, Xavier Maeder

Additive manufacturing offers the opportunity to produce complex geometries from novel alloys with improved properties. Adapting conventional alloys to the process-specific properties can facilitate rapid implementation of these materials in industrial practice. Nevertheless, the processing of conventional alloys by laser powder bed fusion is challenging, particularly in cases of pronounced susceptibility to hot cracking.Previous studies have demonstrated the positive influence of zirconium on the susceptibility to hot cracking of high-strength aluminum alloys, although its influence on precipitation formation, which is immensely important for 2xxx alloys, remains largely unexplored. This work investigates the optimum process parameters and precipitation formation of 2618 modified by zirconium in laser powder bed fusion.The addition of zirconium results in the production of a crack-free, high-density (~99.9%) material. The microstructure is characterized by a trimodal grain size distribution. Very fine (~0.5 mu m) equiaxed grains, nucleated by L12-Al3Zr precipitation at the melt pool boundary, followed by columnar-dendritic grains (5-15 mu m long, 1-3 mu m wide) and coarser equiaxed grains (1-3 mu m) form during solidification. The grain boundaries are populated predominantly by (Al,Cu)9FeNi, but also by Mg2Si, Al2CuMg, and AlCu, which presumably impede grain growth and promote a very fine-grained, low-textured microstructure, even in regions where L12-Al3Zr are absent. The as-built microhardness of 1360 +/- 74 MPa exceeds that of the known high-strength Al alloys tailored to additive manufacturing, AddalloyTM and Scalmalloy (R). The results provide a better understanding of precipitate formation in Zr-modified 2xxx series alloys and pave the way for the commercialization of further 2xxx alloys adapted to additive manufacturing. (c) 2022 The Author(s). Published by Elsevier B.V. CC_BY_NC_ND_4.0

2022

Microprobe analysis of neutron irradiated and autoclaved zirconium niobium claddings using synchrotron-based hard X-ray imaging and spectroscopy

Camelia Borca

Two zirconium alloys (Zr-2.5%Nb) - one oxidized in a pressurized water reactor, the other oxidized in autoclave and used as reference - are analyzed by combining synchrotron-based scanning transmission and fluorescence X-ray microscopy and micro-X-ray absorption spectroscopy (micro-XAS). Two-dimensional zirconium distribution maps recorded on the neutron irradiated and the non-irradiated autoclaved Zr-2.5%Nb alloys clearly allow the localization of the oxide and the metal parts of the interface with a micrometer spatial resolution. Micro-XAS investigations make possible the determination of the speciation of zirconium and niobium both in the oxide and the metal parts of the interface for the irradiated and non-irradiated samples. The coordination environment and/or the valency of zirconium and niobium in the metal and the oxide parts of the interface have been determined for both materials, and interpreted on the basis of comparison with metal and oxide reference compounds. (c) 2008 Elsevier B.V. All rights reserved.