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Surface topography and deuterium retention in polycrystalline hot-rolled W and W–5%Re have been examined after exposure to a low-energy (76 eV), high flux (around 10 22 D/m 2 s) deuterium plasma to an ion fluence of 10 26 D/m 2 at various temperatures. The methods used were confocal laser scanning microscopy and the D( 3 He, p) 4 He nuclear reaction at 3 He energies varied from 0.69 to 4.0 MeV. During exposure to the D plasma at temperatures in the range from 348 to 673 K, small blisters of size in the range from about 1 to about 15 μm, depending on the exposure temperature, are formed on the W and W–5%Re surfaces. In the W–5%Re, the deuterium retention demonstrates its maximum at exposure temperature of 463 K, while in the W this maximum is shifted to 523 K. A difference in the temperature dependence of the D retention for the W and W–5%Re is explained, as a rough approximation, by temperature dependences of the ductility of these materials. [ABSTRACT FROM AUTHOR]

Abstract: Be-seeded, high-flux, deuterium/helium mixture plasma exposure experiments on tungsten target materials have been performed to simulate ITER all tungsten divertor erosion/modification and deposition phenomena. The exposure conditions are kept fixed at a typical low-ion-energy of 60eV and a flux of 3–6×1022/m2/s. Sample temperature is 1123K and plasma exposure times spanning 1050–10,100s are explored. The typical ratio of He/D ions is 0.2 and Be content is 0.2%. A He-induced nanostructure layer is formed on the exposure surfaces of tungsten materials and the surface of the nanostructure is covered by a thin layer of Be and O. A fraction of the re-eroded Be from the target is deposited on a glassy carbon plate with line of sight to the tungsten target. Rutherford backscattering spectrometry analyses show that the Be redeposit layer is in the form of laminae. Small amounts of Mo, W and C are also found in the redeposited Be layer. Elastic recoil detection analyses show that D, He and H are also included in the redeposited Be layer. [Copyright &y& Elsevier]

Abstract: Tungsten materials are candidates for plasma-facing components for the International Thermonuclear Experimental Reactor and the DEMOnstration power plant because of their superior thermophysical properties. Because these materials are not common structural materials like steels, knowledge and strategies to improve the properties are still under development. These strategies discussed here, include new alloying approaches and microstructural stabilization by oxide dispersion strengthened as well as TiC stabilized tungsten based materials. The fracture behavior is improved by using tungsten laminated and tungsten wire reinforced materials. Material development is accompanied by neutron irradiation campaigns. Self-passivation, which is essential in case of loss-of-coolant accidents for plasma facing materials, can be achieved by certain amounts of chromium and titanium. Furthermore, modeling and computer simulation on the influence of alloying elements and heat loading and helium bombardment will be presented. [Copyright &y& Elsevier]

Abstract: The uptake of tritium at 573K by polycrystalline tungsten limiter material pre-exposed to TEXTOR plasmas was investigated by the imaging plate technique (IP) and found to be mostly non-homogeneous on the plasma facing surface. Particularly high concentrations of tritium were apparent in areas attributed to carbon deposition. The surface density of tritium outside of deposition zones was essentially comparable on both sides of the examined tungsten plate. Under a stream of argon at ambient temperature tritium was predominantly released as tritiated water. While tritium is initially liberated with rates in the (MBq/h) range, after a few days the rate drops to about 100Bq/h, decreasing even further thereafter. Under atmospheric conditions the concentration of tritium on the surface remained virtually unchanged over a rather extended period of time, i.e. more than 500d. Tritium in surface zones other than of “deposition” was also firmly trapped at ambient temperature. [Copyright &y& Elsevier]

Abstract: Retention of D in neutron-irradiated W and desorption were examined after plasma exposure at 773K. Deuterium was accumulated at a relatively high concentration up to a large depth of 50–100μm due to the trapping effects of defects uniformly induced in the bulk. A significant D release in a vacuum continued to temperatures⩾1173K because of the small effective diffusion coefficient and the long diffusion distance. Exposure of ion-irradiated W to D2 gas showed a clear correlation between concentrations of trapped and solute D as determined by the trapping–detrapping equilibrium. These observations indicated that the accumulation of tritium in high concentrations is possible even at high temperatures if the concentration of solute tritium is high, and baking at moderate temperatures is ineffective for removal of tritium deeply penetrating into the bulk. Nevertheless, clear enhancement of D release was observed under the presence of solute H. [Copyright &y& Elsevier]