DLC Thin Films and Carbon Nanocomposite Growth by Thermionic Vacuum Arc (TVA) Technology

The aim of this chapter is to report the results on synthesis DLC thin films and carbon nanocomposites by the versatile nanofabrication method based on plasma entitled thermionic vacuum Arc (TVA). TVA technology is based on the localized ignition of the arc plasma in vacuum conditions. Among thin film coating methods by vacuum deposition techniques with high purity, low roughness, and good adhesion on the substrates, TVA is one of the major suitable methods to become a powerful coating technology. Two or three different TVA discharges can be ignited simultaneously in the same chamber for multi-material processing using TVA and separate power supplies. These TVA discharges are localized and do not interfere with each other. Simultaneous two or three TVA discharges were already used for the production of alloy/composite of various materials. This is due to the high versatility concerning the configuration of experimental arrangements, taking into account the number of electron guns, symmetry of the electrodes, relative position of the anode versus cathode, and also the huge opportunity to combine the materials to be deposited: bi- and multi-layers, nanocomposites, or alloys in order to have specific applications. This chapter presents the comparative results concerning the surface-free energy information processing, the reflective index, the hardness, and the morphology to provide a coherent description of the diamond-like carbon films and carbon nanocomposites synthesized by thermionic vacuum arc (TVA) and related configurations where Me = Ag, Al, Cu, Ni, and Ti: binary composites (C-Me, C-Si) and ternary composites (C+Si+Me). The results include reports on the distribution in size, surface, geometry, and dispersion of the nanosized constituents, tailoring and understanding the role of interfaces between structurally or chemically dissimilar phases on bulk properties, as well as the study of physical properties of nanocomposites (structural, chemical, mechanical, tribological). The results presented here could have a great impact on the development of advanced materials and many manufacturing industries, as well as expanding the technologically important field of interface science where the control of the film-substrate interface would be critical

Helium and deuterium irradiation effects in tungsten-based materials with titanium

Pure Tungsten (W) will be used as plasma facing component in fusion devices due to its high melting point, good thermal conductivity and low sputtering yield. However, its structural application as plasma facing component (PFC) is still restricted by its low fracture toughness associated with the high ductile to brittle transition temperature (DBTT). In the present study tungsten titanium (W-Ti) samples were produced by Ti implantation at room temperature and 500 °C with a constant fluence of 2 × 1021 at/m2 and an energy of 100 keV. In order to understand the fundamental mechanisms which govern the behavior of defect dynamics in tungsten under reactor conditions, W-Ti materials were implanted at room temperature with 10 keV of He+ with a constant fluence of 5 × 1021 at/m2 and 5 keV of D+ with fluences in the range of 0.1 × 1021–5 × 1021 at/m2. Surface structure and morphology changes were investigated by scanning electron microscopy and X-ray diffraction. Rutherford backscattering spectrometry, nuclear reaction analysis and thermal desorption spectroscopy methods were used to provide information about the distribution of Ti, He and D on W. No changes in the microstructure were observed after Ti implantation in the W plates. NRA analysis showed that D retention in the W-Ti samples is higher after sequential He and D implantation when compared with single D implantation. The diffractogram of W-Ti samples implanted with He evidence a broadening of the W peaks. This effect is believed to be associated with the high volume fraction of the bubbles that may cause internal stress fields inducing extended defects like dislocations which distort the crystal lattice.info:eu-repo/semantics/publishedVersio

Fuel retention and erosion-deposition on inner wall cladding tiles in JET-ILW

The morphology of beryllium coatings on the Inconel inner wall cladding tiles after JET-ILW campaigns was determined. The focus was on: (i) fuel retention and its share in the overall fuel inventory; (ii) the change of the layer structure and composition. The study is motivated in the view of planned D-T operation in JET. Four tiles were examined: the initial not exposed; one exposed to two campaigns (ILW1-2) and two facing the plasma during ILW1-3. As determined with ion beam and microscopy methods, the initial Be layer (9.0 mu m thick) contained up to 4-5 at.% of impurities, mainly H, O, C, Ni. In the exposed tiles, the impurity content increases to 14-26 at.% (up to 20 at.% O, 1.7 at.% C, 1.0 at.% N, 1.3 at.% Ni and under 0.1 at.% W). The surface composition indicates gettering of O and a long-term retention of N. The Be thickness on the tile exposed to ILW1-2 was between 7.6 and 9.7 mu m, thus indicating erosion in some areas, while the thickness after ILW1-3 increased to 10-12 mu m. The D content was in the range 1.2-3.4x10(17) cm(-2) after ILW1-2 and 3.2-10x10(17) cm(-2) after ILW1-3 on most of the analyzed area, but in the limiter shadow values up to 58 x10(17) cm(-2) were measured. Taking into account the total area of the Be-coated inner wall cladding tiles, the lower limit of D inventory amounts to 5.3x10(22) atoms corresponding to about 176 mg, i.e. somewhat greater than the amount determined on Be limiters. The formation and spalling-off of Be-O particles was revealed

Recent progress in L-H transition studies at JET: Tritium, Helium, Hydrogen and Deuterium

We present an overview of results from a series of L-II transition experiments undertaken at JET since the installation of the ITER-like-wall (JET-ILW), with beryllium wall tiles and a tungsten divertor. Tritium, helium and deuterium plasmas have been investigated. Initial results in tritium show ohmic L-H transitions at low density and the power threshold for the L-H transition (P-LH) is lower in tritium plasmas than in deuterium ones at low densities, while we still lack contrasted data to provide a scaling at high densities. In helium plasmas there is a notable shift of the density at which the power threshold is minimum ((n) over bar (e,min)) to higher values relative to deuterium and hydrogen references. Above (n) over bar (e,min) (He) the L-H power threshold at high densities is similar for D and He plasmas. Transport modelling in slab geometry shows that in helium neoclassical transport competes with interchange-driven transport, unlike in hydrogen isotopes. Measurements of the radial electric field in deuterium plasmas show that E-r shear is not a good indicator of proximity to the L-H transition. Transport analysis of ion heat flux in deuterium plasmas show a non-linearity as density is decreased below (n) over bar (e,min). Lastly, a regression of the JET-ILW deuterium data is compared to the 2008 ITPA scaling law