Gas tungsten arc welding | EPFL Graph Search

Gas tungsten arc welding (GTAW), also known as tungsten inert gas (TIG) welding, is an arc welding process that uses a non-consumable tungsten electrode to produce the weld. The weld area and electrode are protected from oxidation or other atmospheric contamination by an inert shielding gas (argon or helium). A filler metal is normally used, though some welds, known as autogenous welds, or fusion welds do not require it. When helium is used, this is known as heliarc welding. A constant-current welding power supply produces electrical energy, which is conducted across the arc through a column of highly ionized gas and metal vapors known as a plasma. TIG welding is most commonly used to weld thin sections of stainless steel and non-ferrous metals such as aluminum, magnesium, and copper alloys. The process grants the operator greater control over the weld than competing processes such as shielded metal arc welding and gas metal arc welding, allowing stronger, higher-quality welds. Howev

Contribution to the Production and Characterization of W-Y, W-Y2O3 and W-TiC Materials for Fusion Reactors

Lyubomira Veleva

This PhD Thesis work was aimed at investigating the potentiality of tungsten-base materials as structural materials for the future thermonuclear fusion reactors in attempting to develop reduced activation tungsten-base materials with high strength and sufficient ductility, especially in terms of fracture toughness and ductile-to-brittle transition temperature, on the basis of the following ideas: (1) nano-grained materials are expected to show an improved ductility with respect to normal grain-sized materials, and (2) nano-grained materials and materials reinforced with either oxide or carbide particles are expected to show improved strength and radiation resistance, as (i) the particles should act as obstacles for the propagation of mobile dislocations, (ii) the numerous grain boundaries and interfaces between the matrix and the particles should act as sinks for the irradiation-induced defects, and (iii) the particles should also help stabilizing the numerous grain boundaries upon thermal annealing and/or irradiation. A variety of W-base materials, namely W-(0.3-1.0-2.0)Y2O3, W-(0.3-1.0-2.0)Y and W-(0.3-0.9-1.7)TiC materials (in weight percent), have been successfully produced at the laboratory scale by powder metallurgy techniques including mechanical alloying followed by either cold pressing and sintering or hot isostatic pressing (HIPping). X-ray diffractometry and scanning and transmission electron microscopy observations combined with density and microhardness measurements as well as with tensile, Charpy impact, fracture toughness and non-standard three point bending tests allowed the identification of optimal chemical compositions and manufacturing conditions, among those investigated and on the basis of the experimental devices at disposal. Microscopic investigations showed that the materials produced at the laboratory scale are composed of small grains, with sizes between a few tens and a few hundreds nanometers, and contain a high density of small Y2O3 or TiC particles, with sizes between 1 and 50 nm. In the case of W-Y materials, all the Y transformed into Y2O3 during mechanical alloying, due to the high amount of O (around 1 wt.%) present in the milled powders, which is beneficial for reducing the excess O content in the materials. All materials also contain a residual porosity of a few percents, which is typical of materials compacted by HIPping. Macroscopic investigations showed that the HIPped materials exhibit high strength values and a promising resistance to irradiation but poor fracture properties up to a very high temperature of about 1100 °C, due mostly to air contamination of the mechanically alloyed powders and to residual porosity of the compacted materials. The materials produced entirely at the laboratory scale by mechanical alloying and HIPping are of better quality than a W-2Y material that was produced by mechanical alloying at the laboratory scale and compacted by sintering and forging at the Plansee company (Austria). Further studies should focus on reducing the O content in the mechanically alloyed powders and the residual porosity of the compacted materials. The grain size should be also optimized as well as the size distribution, number density and crystallographic features of the oxide and carbide particles.

EPFL2011

Manufacture of ITER feeder sample conductors

Pierluigi Bruzzone, Hongwei Li, Bo Liu, Yu Wu

The ITER feeders are the components that connect the ITER magnet systems located inside the main cryostat to the cryogenics, power-supply and control system interfaces outside the cryostat. The feeder busbars rely on the Cable-In-Conduit Conductor (CICC) design concept as all the conductors for the ITER magnet systems. There are two types of busbars for the feeder systems. One is the Main Busbar (MB) for the TF, CS and PF feeders, and the other is the Corrector Busbar (CB) for the CC feeders. The busbar cable is wound from multiple stage sub-cables made with Cu and superconducting strands. The superconducting material is NbTi for the busbar strands of all feeder systems. All Feeder conductors are provided by China. The R&D programs are needed to acquire knowledge on the behavior of such conductors. Since the conductors are new, some full size copper dummy conductors have been produced for the testing of the cabling parameters, definition of automatic TIG welding of seamless jacket section, elaboration of cable insertion and compaction. Then, two short qualification conductor samples (MB and CB) are prepared in ASIPP, and NbTi advanced strands are produced by Western Superconductor Technology (WST). The details of manufacturing procedures for Feeder conductor samples will be described in this paper. (C) 2013 Elsevier B.V. All rights reserved.

Elsevier Science Sa2013

From materials development to their test in IFMIF: an overview

Nadine Baluc, Nazar Ilchuk, Zbigniew Oksiuta, Robin Schäublin, Philippe Spätig, Minh Quang Tran, Lyubomira Veleva

R&D activities on fusion reactor materials in Switzerland focus on (1) the development of advanced metallic materials for structural applications in plasma-facing (first wall, divertor) and breeding blanket components of the future fusion power reactors, in particular oxide dispersion strengthened reduced activation ferritic steels and tungsten-base materials, (2) the modelling of radiation damage and radiation effects and (3) small specimen test technology for the future International Fusion Materials Irradiation Facility. The main objectives, examples of recent results and future activities are described in the case of these three R&D areas.

2011

Contribution to the Production and Characterization of W-Y, W-Y2O3 and W-TiC Materials for Fusion Reactors

Lyubomira Veleva

EPFL2011