M. Tokitani

Recent progress in understanding the behavior of dust in fusion devices

It has been known for a long time that microscopic dust appears in plasmas in fusion devices. Recently it was shown that dust can be responsible for the termination of long- discharges. Also, in ITER-scale experiments dust can pose safety problems related to its chemical activity, tritium retention and radioactive content. In particular, the presence of dust in the vacuum chamber of ITER is one of the main concerns of the ITER licensing process. Here we review recent progress in the understanding of different experimental and theoretical aspects of the physics of dust dynamics and transport in fusion plasmas and discuss the remaining issues.

Exfoliation of the tungsten fibreform nanostructure by unipolar arcing in the LHD divertor plasma

The tungsten nanostructure (W-fuzz) created in the linear divertor simulator (NAGDIS) was exposed to the Large Helical Device (LHD) divertor plasma for only 2 s (1 shot) to study exfoliation/erosion and microscopic modifications due to the high heat/particle loading under high magnetic field conditions. Very fine and randomly moved unipolar arc trails were clearly observed on about half of the W-fuzz area (6 × 10 mm2). The fuzzy surface was exfoliated by continuously moving arc spots even for the very short exposure time. This is the first observation of unipolar arcing and exfoliation of some areas of the W-fuzz structure itself in a large plasma confinement device with a high magnetic field. The typical width and depth of each arc trail were about 8 µm and 1 µm, respectively, and the arc spots moved randomly on the micrometre scale. The fractality of the arc trails was analysed using a box-counting method, and the fractal dimension (D) of the arc trails was estimated to be D ≈ 1.922. This value indicated that the arc spots moved in Brownian motion, and were scarcely influenced by the magnetic field. One should note that such a large scale exfoliation due to unipolar arcing may enhance the surface erosion of the tungsten armour and act as a serious impurity source for fusion plasmas.

Extended steady-state and high-beta regimes of net-current free heliotron plasmas in the Large Helical Device

The performance of net-current free heliotron plasmas has been developed by findings of innovative operational scenarios in conjunction with an upgrade of the heating power and the pumping/fuelling capability in the Large Helical Device (LHD). Consequently, the operational regime has been extended, in particular, with regard to high density, long pulse length and high beta. Diversified studies in LHD have elucidated the advantages of net-current free heliotron plasmas. In particular, an internal diffusion barrier (IDB) by a combination of efficient pumping of the local island divertor function and core fuelling by pellet injection has realized a super dense core as high as 5 × 10 20 m -3 , which stimulates an attractive super dense core reactor. Achievements of a volume averaged beta of 4.5% and a discharge duration of 54 min with a total input energy of 1.6 GJ (490 kW on average) are also highlighted. The progress of LHD experiments in these two years is overviewed by highlighting IDB, high β and long pulse.

High-density plasma with internal diffusion barrier in the Large Helical Device

An attractive high-density operational regime which is a so-called internal diffusion barrier (IDB) has been discovered in a helical divertor configuration on the Large Helical Device (LHD). The IDB is characterized by steep density gradients and the plasma profile is divided by the IDB into a high-density core plasma and a low density mantle plasma. The IDB enables the core plasma to access the high-density/high-pressure regime. The attainable central density exceeds 1 × 1021 m−3 and the central pressure reaches ≈1.5 times atmospheric pressure. Core pellet fuelling is absolutely essential for the IDB formation and it is reproducibly obtained by employing intensive multiple-pellet injection. In the IDB core plasma, the particle diffusion coefficient is kept at a considerably low level, 0.05 m2 s−1, in spite of high-density and steep-density gradients whereas an inward particle convection velocity is not observed.

Microscopic modification of wall surface by glow discharge cleaning and its impact on vacuum properties of LHD

Glow discharge cleaning (GDC) is a widely used technique for wall conditioning in fusion experimental devices. Though the cleaning effects of GDC are essentially related to the microscopic modification of the wall surface, there are few reports about it. In the present study, specimens of wall materials were exposed to GDC plasma of helium, hydrogen and neon in the Large Helical Device (LHD) by using the retractable material probe transfer system and irradiation damage was examined by transmission electron microscopy to understand the underlying microscopic mechanism of GDC. In the case of Ne-GDC, the specimen surface was covered with a thick deposited layer of Fe and Cr but no radiation induced defects were observed. Due to the high sputtering efficiency and very shallow penetration, it is likely that neon atoms effectively sputter the surface contamination without leaving serious damage or remaining in the sub-surface region. After the Ne-GDC phase, retained Ne can be successfully removed by the following short hydrogen GDC. It was shown that a two-step GDC with Ne and H is very effective to clean the metallic surface of the LHD.

Steady-state operation using a dipole mode ion cyclotron heating antenna and 77 GHz electron cyclotron heating in the Large Helical Device

The steady-state operation of high-performance plasmas in the Large Helical Device (LHD) has progressed since the 2010 IAEA Conference in Korea by means of a newly installed ion cyclotron heating (ICH) antenna (HAS antenna) and an improved electron cyclotron heating (ECH) system. The HAS antenna can control the launched parallel wave number and heat the core plasma efficiently in the case of dipole mode operation. Understanding of the physics and technology of wave heating, particle and heat flow balances, and plasma?wall interactions in LHD has also improved. The heating power of steady-state ICH and ECH exceeded 1?MW and 500?kW, respectively, and a higher density helium plasma with minority hydrogen ions was maintained using the HAS antenna and new 77?GHz gyrotrons. As a result, plasma performance improved, e.g. electron temperature of more than 2?keV at a density of more than 2???1019?m?3 became possible for more than 1?min. Heat flow balance and particle flux balance of steady-state operation were evaluated. Particle balance analysis indicated that externally fed helium and hydrogen particles were mainly absorbed by the chamber wall and divertor plates, even after the 54?min operation.

In situ measurement of hydrogen isotope retention using a high heat flux plasma generator with ion beam analysis

We present Plasma Surface Dynamics with Ion Beam Analysis (PS-DIBA), a device that makes it possible to carry out in-situ ion beam analysis of deuterium retention under high-density plasma irradiation, relevant to divertor plasma conditions. The device is equipped with a newly developed novel dc plasma source, which can generate high-density plasma (>1018 m−3) with an electron temperature of around 5 eV in steady state. The dynamic behavior of deuterium retention in ITER R&D tungsten exposed to deuterium plasma was observed by using nuclear reaction analysis in PS-DIBA. The results indicate that deuterium retention strongly depends on the surface temperature of the tungsten sample during plasma exposure. After plasma termination, the retained deuterium in tungsten decreases faster than in isotropic graphite.

Effect of a magnetic island on the three-dimensional structure of edge radiation and its consequences on detachment in the Large Helical Device (EX-D)

In this paper the effects of an externally produced magnetic island (MI) on the 3D radiation structure in attached and detached plasmas as predicted by the EMC3-EIRENE code are clearly seen in the imaging bolometer (IRVB) data from two different views of the plasma, experimentally confirming the role that the MI plays in the detachment process. With the addition of the MI the carbon radiation profile from the code in a poloidal cross-section becomes more localized near the helical divertor x-points (HDXs). This is reflected in the focusing of the radiation patterns corresponding to the HDX in both the IRVB and code data in images corresponding to the IRVB field of view (FOV). Detachment results in a more asymmetric radiation profile in the poloidal cross-section code data with localized peaks near the HDX and magnetic island x-points (MIXs). The radiation from the MIXs is reflected in strong radiation from the corresponding location in the IRVB FOV from both code and IRVB data. Also the change in the position of the MI results in a change in the position of the localized radiation peak as predicted by the code. However, the relative increase in the radiation from the MIXs is greater in the code data than in the IRVB data for reasons which are so far unknown. Also similar discharges show detachment with the MI, albeit at a lower density than the discharge without the MI. This work confirms the previous conclusions that the MI enhances the localization of the radiation and is conducive to achieving and sustaining the detachment.

Comparative divertor-transport study for helical devices

Using the island divertors (IDs) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine sizes following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. The fundamental role of low-order magnetic islands in both divertor concepts is emphasized. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime, which is absent from W7-AS and LHD, is predicted to exist for W7-X. The paper focuses on identifying and understanding the role of divertors for high density plasma operations in helical devices. In this regard, special attention is paid to investigating the divertor function for controlling intrinsic impurities. Impurity transport behaviour and wall-sputtering processes of CX-neutrals are studied under different divertor plasma conditions. A divertor retention effect on intrinsic impurities at high SOL collisonalities is predicted for all the three devices. The required SOL plasma conditions and the underlying mechanisms are analysed in detail. Numerical results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. Different SOL transport regimes are numerically identified for the standard divertor configuration of W7-X and the possible consequences on high density plasmas are assessed. All the EMC3-EIRENE simulations presented in this paper are based on vacuum fields and comparisons with local diagnostics are made for low-s plasmas.

Studies of dust transport in long pulse plasma discharges in the large helical device

Three-dimensional trajectories of incandescent dust particles in plasmas were observed with stereoscopic fast framing cameras in a large helical device. It proved that the dust is located in the peripheral plasma and most of the dust moves along the magnetic field lines with acceleration in the direction that corresponds to the plasma flow. ICRF heated long pulse plasma discharges were terminated with the release of large amounts of dust from a closed divertor region. After the experimental campaign, the traces of exfoliation of carbon rich mixed-material deposition layers were found in the divertor region. Transport of carbon dust is investigated using a modified dust transport simulation code, which can explain the observed dust trajectories. It also shows that controlling the radius of the dust particles to less than 1 mm is necessary to prevent the plasma termination by penetration of dust for the long pulse discharges. Dust transport simulation including heavy metal dust particles demonstrates that high heating power operation is effective for shielding the main plasma from dust penetration by an enhanced plasma flow effect and a high heat load onto the dust particles in the peripheral plasma. It shows a more powerful penetration characteristic of tungsten dust particles compared to that of carbon and iron dust particles.