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Helium (He) retention in re-crystallized tungsten (W) irradiated with He+ ions at elevated temperatures (700–1200 K) and fluences in the range of 1020–1022 He/m2 was investigated by means of thermal desorption spectroscopy (TDS). Corresponding surface modifications were analyzed by scanning electron microscopy. Blisters were observed after irradiation at 700 and 1000 K, while the increase of the irradiation temperature up to 1200 K led to development of a complex sponge-like structure on the W surface. Significant surface transformations correlated with appearance of low temperature peaks in TDS spectra below the irradiation temperature. Possible mechanisms and explanations are discussed. [ABSTRACT FROM AUTHOR]

Helium (He) isotope exchange in tungsten (W) during sequential irradiation by 3 keV 4He and 3He ions at room (RT) and elevated temperatures (700-1200 K) was investigated. The total He fluence was in the range of 5 × 1021–1.8 × 1022 He/m2 to provide a saturation of the surface layer. The amount of He retained in W after irradiation was measured using in-situ (up to 1500 K) and ex-situ (up to 2500 K) thermal desorption spectroscopy (TDS). Air exposure influenced TDS spectra, but clear were no noticeable differences between 3He and 4He TDS spectra after single irradiation. Subsequent irradiation with different He isotopes demonstrated a very efficient isotope exchange already at room temperature. The substitution of He atoms goes even faster with increasing irradiation temperature. Mechanisms of observed processes are discussed on the base of a simple model. [ABSTRACT FROM AUTHOR]

The fully reduced activation brazing alloy TiZr4Be and a tantalum intermediate layer are considered to be used for joining tungsten (W) to reduced-activation ferritic-martensitic steel (RAFM) for future DEMO reactor application. Deuterium retention in the W-Rusfer joints and separate elements was investigated with the focus on the intermediate brazing layer. The samples were exposed in deuterium gas (p = 1–104 Pa, T = 300–600°C) and plasma discharge (T = 600°C). An acceptable deuterium concentration was observed after gas exposure at the pressure of 1 Pa, that is relevant to operating conditions of future fusion devices, however the brazing alloy and tantalum accumulate a large amount of deuterium in the case of the pressure above 100 Pa that leads to failure of the joint. The D retention after D plasma irradiation was substantial, but nearly independent on the fluence and the thickness of the W layer due to absorption from surrounding D 2 gas. [ABSTRACT FROM AUTHOR]

Abstract: Accumulation and recovery of radiation defects under/after self-ion irradiation in tungsten (W) have been investigated via decoration with deuterium (D) and scanning transmission electron microscopy (STEM). The deuterium was incorporated in damaged material by low-energy D plasma. The D concentration at radiation-induced defects in each sample was subsequently measured by nuclear reaction analysis allowing determination of the D concentration at depths up to 6μm. The total D retention was measured by thermal desorption spectroscopy. It was shown that pre-irradiation with self-ions led to rather high D concentration (⩾ 0.1at.%) in W even at high temperatures (⩾ 700K) due to formation of defects with high de-trapping energy for deuterium. The annealing of defects with low trapping energy for D occurs intensively in the temperature range between 300 and 700K. The radiation-induced defects with high de-trapping energy for D are thermally stable at least up to 1100K. The rearrangement and partial healing of dislocations as well as coalescence of small clusters in a big ones accompanied by a reduction of the total density of defects was observed by STEM after annealing of radiation-induced defects in recrystallized tungsten at 1000K. The D retention monotonically decreases in recrystallized W with increasing of annealing temperature up to 1100K that is in agreement with the reduction of radiation defect density observed by STEM. However, an increase of the D retention in ‘as received’ W pre-irradiated with self-ions at annealing temperature of around 1000K was found. The increase of the D retention at annealing temperature of ∼1000K was not observed in the case of recrystallized W pre-irradiated with self-ions. The mechanism of recovery of radiation-induced defects in dependence on the initial intrinsic defects (grain size, impurities, etc.) in W is discussed. [Copyright &y& Elsevier]

Abstract: Deuterium retention in C–D and C–W–D mixed films deposited in a magnetron discharge in deuterium was studied. The deuterium content in the C–D films was in the range D/C=0.65–0.75 for grounded substrates and 0.45–0.6 if a bias of −60V was applied. The deuterium content in the C–W–D films was in the range of D/(C+W)=0.2–0.4 without strong dependence on the C/W ratio (in the range of 0.7–10) and the substrate potential (in the range from 0 to −60V). Deuterium release from the C–W–D films was observed at lower temperatures than that from the C–D films. [Copyright &y& Elsevier]

Abstract Positron annihilation lifetime spectroscopy (PALS) was applied to study the annealing of radiation-induced defects in polycrystalline tungsten (W) irradiated with 21.6 MeV protons at 100 °C up to a fluence of 5 × 1015 p/cm2. Three components were observed in the measured spectra: short-lifetime of 100–120 ps (positron annihilation in the defect-free W lattice), medium-lifetime of ∼190–330 ps (annihilation at mono-vacancies and small vacancy cluster containing ∼ 2–4 vacancies) and long-lifetime of ∼500 ps (annihilation in large vacancy clusters containing more than 10 vacancies). The irradiation of W with protons at 100 °C, primary, led to the formation of mono-vacancies, self-interstitial defects were created as well but migrated towards sinks during the irradiation. Onset of vacancy diffusion in W starts already at 200 °C before defect recovery stage III. After annealing at ∼400 °C, a sharp drop in the intensity of the positron medium-life component together with a simultaneous increase in positron lifetime from ∼220 to ∼280 ps is observed, and a long-life component appears. This indicates migration and annealing of vacancies and their agglomeration in large vacancy clusters. After annealing at 500–700C, the intensity of long-life component increases indicating the growth of large vacancy clusters but at 900 °C they anneal completely as the mean lifetime recovers nearly to the value measured in the un-irradiated material. [ABSTRACT FROM AUTHOR]

A systematic and attentive comparison of the deuterium (D) retention in tungsten (W) coatings with different nano-crystalline structures after the plasma exposure in comparison with polycrystalline tungsten (PCW) is presented. While a wide database is available for PCW, only a few data about the D retention in coatings with different structures exist. The D retention in W coatings produced by three different deposition techniques on different substrates was studied with respect to the influence of (a) coating crystallite size, (b) coating thickness, (c) specimen temperature during D plasma exposure, (d) presence of argon (Ar) used as working gas during the coating deposition and (e) substrate material. It is shown that the variation of the processing parameters, such as temperature, deposition rate, Ar implantation, etc. even within one deposition method results in different grain size distributions and structure of coating and has a significant effect on the D retention. It is revealed that the substrate material and the presence of Ar in a coating play a minor role in the D retention in the coating. It is shown that both the D concentration and the D retention in coatings drastically increase with decreasing the grain size. Consequently, in the case of using of W coating as a protective layer of a structural material, a compromise in the development of nanostructured tungsten films is necessary to keep the hydrogen isotope concentration at an acceptable level. [ABSTRACT FROM AUTHOR]

Using a beam-plasma discharge device operating on helium (He), tungsten with the fuzz on the surface (W f) has been formed by irradiating polycrystalline tungsten (W) samples with He ions with an energy of ∼150 eV and the fluence of ∼6 × 1024 He/m2 at the temperature of 1273 K. The deuterium (D) retention in W f annealed at different temperatures was studied by thermal desorption spectroscopy (TDS). Before and after annealing at temperatures of 1000,1200,1400 and 1600 K, W f was irradiated at room temperature by 2 keV D 3 +(667 eV/D) ions with the fluence of 1021 D/m2, then in-situ TDS was performed after each irradiation. Annealing W above 1200 K clearly changes the retention mechanism of D: the TDS spectrum consisting of multiple peaks changes to an almost single-peak spectrum. Annealing at 1600 K leads to surface smoothing and the decrease of the D retention by a factor of two compared to the annealing at 1000 K. This can be explained by an increase of the reflection coefficient for the flat W surface. However, the D retention in W f is significantly higher compared to that in W without He plasma exposure even after annealing at 1600 K, because there are still He bubbles in W f that effectively trap D [ABSTRACT FROM AUTHOR]

In order to simulate ITER transient events with surface heat load parameters relevant to edge-localized-mode (ELM) impacts, tungsten samples were exposed to pulsed heat loads using pure deuterium (D) and with 10% helium (He) seeding plasmas in quasi-stationary high-current plasma gun QSPA-T. The pulse duration was 1 ms that is relevant to ELMs and number of pulses was varied from one to thirty. The power load was 0.7 MJ/m2 that is below the tungsten melting temperature (T m). Two tungsten samples were used, namely, polycrystalline tungsten without (W) and with pre-existing He-induced W 'fuzz' (W f). Similar to the steady state plasma, the presence of He in tungsten leads to a reduction of the D retention during transient events at temperature below T m. We explained it as interruption of the D diffusion towards to the bulk of tungsten by (i) the strain field induced by He bubbles and (ii) the formation of interconnected He bubbles at high temperature which leads to an open porosity for accelerated D desorption, thus, decreasing the D influx into the tungsten bulk. But as the He retention in W f decreases below 1019 He/m2, the effect of He on the D retention after the plasma gun irradiation disappears: the D retention in W and W f is the same after 30 pulses of the exposure to pure D plasma. In both cases of pure D and He seeded D plasma gun exposures, the D retention is higher compared to the steady state plasma exposure at sample temperature above 600 K. [ABSTRACT FROM AUTHOR]

The effect of He-induced defects in tungsten on the efficiency of trapping of deuterium ions in the subsurface layer was studied using thermal desorption spectroscopy (TDS). The W sample was pre-irradiated with 3 keV helium ions at room temperature and various fluences in the range of 1019 – 5 × 1021 He/m2. Then, it was exposed to a probe fluence of 1019 D/m2 of 2 keV D 3 + (670 eV/D) ions, and in-situ TDS was performed. The de-trapping energy for D atoms increased with the increase of the He pre-irradiation fluence. On the other hand, a strong decrease in the D retention was observed if the He fluence increased above 1021 He/m2. At the highest He fluence of 5 × 1021 He/m2 deuterium trapping was possible only after partial release of He atoms. By comparison of experimental TDS spectra with modeling, the de-trapping energies of D atoms from various defects were estimated. [ABSTRACT FROM AUTHOR]

• The entire surface of gaps was covered by Li evaporated and splashed from Li limiter. • Heavy carbide precipitates were found deep in narrow gaps. • Deuterium and helium are accumulated in the tiles due to ion implantation. • Mainly regular but not cleaning discharges were responsible for effects in gaps. • High frequency auto-oscillating discharges were suggested to play a role in processes in gaps. Surfaces facing the gap between W tiles of the ring limiter of tokamak T-10 were analyzed after T-10 decommissioning using LIBS, SEM/EDA, XRD, TDS, and NRA techniques. Gaps with the width of 5 mm and 0.1 mm were nearly completely covered to their full depths of 22 and 15 mm, respectively, by a deposited film. The film was formed mainly by deposition of lithium that came from Li limiter and transformed in air to Li 2 CO 3 and Li 2 O. Carbon was deposited from volatile hydrocarbons sputtered from the tokamak walls. Besides, carbon appeared due to chemical reaction with lithium in air. Chemical interactions of W with C, O, and Li led to formation W 2 C, WC, WO 2 , and Li 2 WO 4. Carbides formed in W over the entire surface to the full depth of the gaps. Trapping of deuterium and helium in tiles was demonstrated. Possible influence of auto-oscillating discharges on ionization and ion trapping of C,D, and He in gaps is discussed. [ABSTRACT FROM AUTHOR]