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• T permeability was obtained in tungsten in W via a gas-driven permeation technique with a flowing H-T permeation system. • T permeability was measured in low T partial pressure of(5.8-6.8)x10-2 Pa over the temperature range from 573 to 873 K. • A scanning Auger microscope was used to characterize surface conditions of W before and after T permeation experiment. A flowing hydrogen-tritium permeation system was used to investigate tritium permeation behavior in polycrystalline tungsten. Tritium permeability was measured with T 2 partial pressure ranging from 5.8 × 10−2 Pa to 6.8 × 10−2 Pa over the temperature range from 573 to 873 K. The obtained tritium permeability in polycrystalline tungsten was P T [mol m−1s−1 Pa−0.5] = (2.13 × 10−9)exp(−93.32 [kJ mol−1]/ RT). This is one of the first tritium permeability data from any polycrystalline tungsten via a gas-driven permeation technique. The tritium permeability in polycrystalline tungsten was compared with the reported hydrogen isotope permeability in other polycrystalline tungsten. [ABSTRACT FROM AUTHOR]

The present work is a continuation of a recent research to develop and optimize positron annihilation spectroscopy (PAS) for characterizing neutron-irradiated tungsten. Tungsten samples were exposed to neutrons in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory and damaged to 0.025 and 0.3 dpa. Subsequently, they were exposed to deuterium plasmas in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory. The implanted deuterium was desorbed through sample heating to 900 °C, and Doppler broadening (DB)-PAS was performed both before and after heating. Results show that deuterium impregnated tungsten is identified as having a smaller S-parameter. The S-parameter increases after deuterium desorption. Microstructural changes also occur during sample heating. These effects can be isolated from deuterium desorption by comparing the S-parameters from the deuterium-free back face with the deuterium-implanted front face. The application of using DB-PAS to examine deuterium retention in tungsten is examined. [ABSTRACT FROM AUTHOR]

Abstract: Recent progress in experiments, modeling and theory has increased confidence that ITER can achieve its goal of high Q (>10) in inductive operation. Further, experimental results in the ‘improved hybrid’ regime suggest a possibility of high Q (>10) operation in a long pulse (>1000s) with benign ELMs. However, considerable uncertainty still exists in the prediction of several key aspects; for example, tritium retention, disruption, impurity control, ELMs, SOL transport and dust. ITER requires flexibility of operation scenarios at least in the early operation phase to accommodate uncertainties in prediction, to explore wide operational spaces and to incorporate newly developed control schemes. Because of these uncertainties conservative assumptions are adopted in performance predictions, and step-wise implementation of reactor-relevant plasma-facing material, such as tungsten, are planned. [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]

One of the ITER diagnostics having reached an advanced design stage is a diagnostic RGA for the divertor, i.e. residual gas analysis system for the ITER divertor, which is intended to sample the divertor pumping duct region during the plasma pulse and to have a response time compatible with plasma particle and impurity lifetimes in the divertor region. Main emphasis is placed on helium (He) concentration in the ducts, as well as the relative concentration between the hydrogen isotopes (mainly in the form of diatomic molecules of H, D, and T). Measurement of the concentration of radiative gases, such as neon (Ne) and nitrogen (N 2 ), is also intended. Numerical modeling of the gas flow from the sampled region to the cluster of analysis sensors, through a long (∼8 m long, ∼110 mm diameter) sampling pipe originating from a pressure reducing orifice, confirm that the desired response time (∼1 s for He or D 2 ) is achieved with the present design. [ABSTRACT FROM AUTHOR]

Abstract: Precipitation in subsurface bubbles is a key process that governs how hydrogen isotopes migrate through and become trapped within plasma-exposed tungsten. We describe a continuum-scale model of hydrogen diffusion in plasma-exposed materials that includes the effects of precipitation. The model can account for bubble expansion via dislocation loop punching, using an accurate equation of state to determine the internal pressure. This information is used to predict amount of hydrogen trapped by bubbles, as well as the conditions where the bubbles become saturated. In an effort to validate the underlying assumptions, we compare our results with published positron annihilation and thermal desorption spectroscopy data, as well as our own measurements using the tritium plasma experiment (TPE). [Copyright &y& Elsevier]