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A tungsten (W) deposition layer is formed while a sample is exposed to a helium (He) plasma at the surface temperature of 473 or 573 K. The formed He-W co-deposition layer was composed of fine grains, the size of which increased with the surface temperature. The samples with the co-deposition layer was exposed to a deuterium (D2) plasma, and the D retention was investigated using thermal desorption spectroscopy. It was found that the co-deposition layer has quite different D retention characteristics compared with bulk W.

Rhodium and ruthenium thin film coatings were conducted on tungsten samples by using a magnetron sputtering device; then, we irradiated He plasma to the samples in the linear plasma device NAGDIS-II (Nagoya Divertor Simulator). Fuzzy nanostructures were formed on rhodium and ruthenium samples when the surface temperatures were ∼950 and ∼1200 K, respectively. When the surface temperature was high, i.e., >1200 K, it was found that tungsten atoms diffused across the rhodium film and reached the film surface, and tungsten-fuzz was formed over the rhodium layer. From TEM analysis of ruthenium fibers, it was identified that there were thin parts in the fibers, and some fibers had no He bubbles. It was likely that fiber growth mechanism on ruthenium was different from the other metals.

We have developed a compact divertor plasma simulator (CDPS) that can produce steady-state deuterium and/or helium plasmas with densities above ∼ 1018 m−3 for Plasma-Wall Interaction (PWI) studies of neutron-irradiated materials. The maximum particle flux is about 1022 m−2s−1. The CDPS was installed and is being operated in the radiation-controlled area of the International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University. We are able to control sample temperature within uncertainty of 5 °C during plasma exposure by adjusting the cooling air flow rate to the sample holder. The CDPS has a sample-carrier system, which makes it possible to transfer a plasma-irradiated sample from the sample holder to an infrared heater for analysis using thermal desorption spectroscopy (TDS) without exposing it to the air. This avoids the oxidation of the sample and minimizes the time between the end of plasma exposure and TDS analysis. An ITER-like tungsten (W) sample (A.L.M.T. Corp.), which has been irradiated by neutrons to 0.06 dpa in a fission reactor, was exposed to a deuterium plasma in the CDPS. The experimental results clearly show that the total deuterium retention in the neutron-irradiated W sample increases significantly in comparison with a pristine W, as demonstrated by broadening of the TDS spectrum at high temperatures.

A fiberform nanostructured layer is formed on a tungsten (W) surface using helium (He) plasma irradiation. The behavior of the nanostructures and He bubbles were observed during annealing via in situ transmission electron microscopy. Notable changes in the nanostructures occurred at 1223 K. A collapse of the nanostructures was observed along with a collapse of the He bubbles during annealing.

Nanostructured tungsten formed by the exposure to helium plasma in a linear plasma device was installed in the large helical device (LHD). After the exposure in a series of experiments in the 2012 fiscal year campaign in LHD, the samples were analyzed by scanning electron microscope (SEM), transmission electron microscope (TEM), and energy dispersion x-ray spectroscopy (EDX). It was found that part of the nanostructures was totally covered with carbon based material probably from divertor, while some other parts were eroded by sputtering. On the erosion dominant region, it was revealed that the head part of nanostructures was sputtered and the surface became rounded, but the nanostructures still remained near the surface. Optical reflectance of the material was measured, and it was found that the morphology changes increased the optical reflectivity up to ∼10% from typically less than 1%. The possibility and limitation of the nanostructured tungsten as a light absorber (viewing dump) are discussed.

First mirrors in fusion devices have to retain the reflectivity and the degradation mechanism and dominant plasma operations that affect the reflectivity have to be understood. The reflectivity of visible laser light (635 nm) of a corner cube mirror is almost entirely determined by the initial hydrogen glow discharges for wall conditioning before conducting a main plasma experiment in the Large Helical Devices (LHD). The hydrogen glow discharge forms a carbon deposition layer on the mirror surface, which degrades its reflectivity. A neon glow discharge and subsequent main discharges had less effect.

Deuterium retention and trapping in tungsten plasma sprayed layers after low energy plasma irradiation was investigated by means of thermal desorption technique. Plasma irradiation at low temperatures ( 650K) showed significant increase of the retention in the plasma sprayed layers contrary to the polycrystalline material. Thermal desorption spectra showed extremely high temperatures of the peak positions which are non-typical for the tungsten materials. The possible reasons for such high deuterium retention in the plasma sprayed layers after irradiation at high temperatures and the possible trapping mechanisms are discussed.

The key to improving the heat load of vacuum plasma sprayed tungsten coatings on low activation ferritic steel maintained at low temperatures is elimination of stratified low-density layers with many large pores, in which thermal cracks propagate preferentially. The low-density layers are formed owing to the deposition of large solidified tungsten particles, which remain mainly at the periphery of the spray stream. In this study, by shading this periphery, partially homogeneous tungsten coatings without large pores were successfully obtained. The coatings are expected to show good heat load, which is feasible for nuclear fusion applications.

Deuterium retention after low energy plasma exposure was investigated using the polycrystalline tungsten samples of which grain elongation directions are parallel and perpendicular with respect to the surface. The deuterium retention calculated from TDS spectra is 2-10 times larger in the sample with perpendicular grain elongation direction than in the parallel one for the irradiation temperatures of 450-750K. Removing of the grain elongation anisotropy by the recrystallization resulted in disappearance of the difference in the deuterium retention of the both kinds of samples. No bubbles and no blisters have been observed in the near surface layer of the samples. Plasma irradiation should produce high energy defects that are responsible for the deuterium retention. The difference in the retention of the both kinds of samples seems to be attributed to the different effective diffusion coefficients depending on the grain elongation direction. c

The ion cyclotron wall conditioning (ICWC) is one of the conditioning methods to reduce impurities and to remove tritium from the plasma facing components. Among the advantages of ICWC are the possible operation under strong magnetic field for fully torus area based on the charge exchange damage observed in thin SS samples arranged on a hexahxedron block holder with three different facings, the areas influenced by ICWC is estimated. On the plasma facing area of the material holder, high density of helium bubbles is observed by transmission electron microscope (TEM). But the other areas show no observable damage. The fact that the bubble were observed only in a sample facing the plasma implies that the effective particles, most probably charge exchange neutrals come to the wall straightly Thus, cleaning of the surfaces un-exposed to plasma directly and those in shadow area is difficult by ICWC.

The neutral beam injection facility in the National Institute of Advanced Industrial Science and Technology was used to irradiate a polycrystalline tungsten specimen with high energy and high flux helium and hydrogen particles. The incidence energy and flux of the beam shot were 25 keV and 8.8 × 1022 particles/m2 s, respectively. The duration of each shot was approximately 30 ms, with 6 min intervals between each shot. Surface temperatures over 1800 K were attained. In the two cases of helium irradiation, total fluence of either 1.5 × 1022 He/m2 or 4.0 × 1022 He/m2 was selected. In the former case, large sized blisters with diameter of 500 nm were densely observed. While, the latter case, the blisters were disappeared and fine nanobranch structures appeared instead. Cross-sectional observations using a transmission electron microscope (TEM) with the focused ion beam (FIB) technique were performed. According to the TEM image, after irradiation with a beam shot of total fluence 4.0 × 1022 He/m2 , there were very dense fine helium bubbles in the tungsten of sizes 1-50 nm. As the helium bubbles grew the density of the tungsten matrix drastically decreased as a result of void swelling. These effects were not seen in hydrogen irradiation case.

Various phenomena of plasma-wall interactions during long duration discharges in TRIAM-1M are investigated from macroscopic and microscopic viewpoints. It is found that the density dependence of the hydrogen neutral flux decay length is not very sensitive to the density (i.e. ne−0.2 ). Neutral transport through the scrape-off layer is important for structural formation of hydrogen recycling. The hydrogen retention in the co-deposited layer, which was obtained using a material probe during long duration discharges, is consistent with the global wall pumping rate estimated from particle balance analysis. The wall temperature and co-deposition play important roles on hydrogen re-emission and hydrogen absorption, respectively. Oxygen impurities should affect the erosion and deposition properties of the plasma-facing wall as well as the hydrogen retention property of the co-deposited layer. The co-deposition of hydrogen with molybdenum and the oxygen impurities both play important roles in the hydrogen recycling.

A high-power-density ion beam system with high-repetition pulses was successfully developed. In the ITER (International Thermonuclear Experimental Reactor), it is anticipated that an intermittent thermal flux, due to the edge localized mode (ELM), to the plasma facing materials causes severe damage of the mechanical properties. Therefore, it is very important to study the effect of ELM phenomena. We already developed an ion beam system with a power density as high as ∼1 GW/m2 around the focal point of the beam. In order to imitate the intermittent high-power-density pulsed flux, we modified the beam operation method and part of the acceleration power supply. A pulsed helium ion beam with the beam width of 2 ms and 4 ms intervals between pulses was successfully extracted. In this case, beam energy, current and power were ∼22 keV, ∼40 A, and ∼0.88 MW, respectively. This high-repetition pulsed helium ion beam with high power density (∼300 MW/m2 ) was irradiated to a tungsten material. It was found that this repetitive short-pulse irradiation caused less surface damage compared with long-pulse irradiation, even when the total amount of irradiation fluence (1.5× 1022 particles/m2) was the same for each condition. This would provide important data for the design of ITER diverter.

QUEST focuses on the steady state operation of the spherical tokamak by controlled PWI and electron Bernstein wave current drive. One of the main purposes of QUEST is an achievement of long duration discharge with MW-class injected power. As the result, QUEST should be operated in the challenging region on heat and particle handling. To do the particle handling, high temperature all metal wall up to 623 K and closed divertors are planned, which is to realize the steady-state operation under recycling ratio, R = 1. This is a dispensable check to DEMO, because wall pumping should be avoided as possible in the view of tritium retention. The QUEST project will be developed in increment step such as, I. low β steady state operation in limiter configuration, II. low β steady state operation in divertor configuration, III. relatively high β steady state operation in closed divertor configuration. Phase I in the project corresponds to these two years, and final goal of phase I is to make full current drive plasma up to 20 kA. Closed divertor will be designed and tested in the Phase II. QUEST is running from Oct., 2008 and the first results are introduced.