[en] A compact and powerful plasma generator was developed with the use of directly heated lanthanum hexaboride cathode which is composed of four units, convenient to handle, and designed to absorb thermal stress. A high plasma density of around 1 x 10¹⁹ m^-3 was generated with a diameter of more than 10 cm in a cylindrical discharge chamber, appropriate for simulation studies on edge plasma physics. (author)

[en] The usage of the numerical system 'LHD LINES OF FORCE.EXE' is described in detail here. The primary purpose of the numerical system 'LHD LINES OF FORCE.EXE' is to analyze interference of lines of force with various structures installed in the vacuum vessel of the LHD. The second purpose of the numerical system 'LHD LINES OF FORCE.EXE' is to show the structure of lines of force and the magnetic surface in front of plasma measurement devices. The feature of the numerical system 'LHD LINES OF FORCE.EXE' is to be able to analyze the interference for arbitrary being installed in the LHD. A GUI (Graphical User Interface) environment is prepared for the preparation of a lot of parameters to run the numerical system 'LHD LINES OF FORCE.EXE'. The numerical results are output by a PostScript file (ps-file). A GUI is also prepared to reduce the file size of the numerical results by converting the ps-file into png (Portable Network Graphics) files. (author)

[en] Linear device NAGDIS-I (Nagoya University Divertor Simulator) with high heat flux plasma has been constructed for simulation studies on plasma-surface interactions in nuclear fusion reactor. High heat flux plasma is generated by PIG discharge using a sophisticated LaB₆ hot cathode assembly. The plasma has a high electron density (above 1 x 10¹⁹m^-3 in a pulse discharge, 4 x 10¹⁸m^-3 in steady state), a large diameter (12 cm) and a long column (above 2m). For the sake of parameter measurement of the high heat flux plasma, fast scanning probe system with high spatial resolution (2 mm) and wide measurement range (20 cm in stroke) has been developed. The details of NAGDIS-I are described, concentrating on the cathode structure, the fast scanning probe system, the plasma parameters and the heat flux. (author)

[en] Edge magnetic structures in helical devices are complicated in comparison with those in poloidal divertor tokamaks, resulting edge transport properties unique to helical devices. The results of experimental observations and computer simulations of the Large Helical Device and Wendelstein7-AS are summarized ; they suggest that cross-field transport plays important role in reduction of plasma pressure along the field lines. (author)

[en] A simple new tool for deposition layer studies, a directional material probe, is proposed. The probe, which examines the directionality of deposition layer formation, consists of a flat disk and pin. If deposits have directionality, a shadow of the pin is formed on the deposition layer on the disk. If no shadow appears on the deposition layer, this suggests that the deposition layer was formed isotropically. The probe can be applied to plasma-wall interaction studies in fusion devices and laboratory plasma devices such as linear divertor simulators to reveal the material migration mechanisms in such devices. The directional material probe method has been applied to plasma-wall interaction studies in the Large Helical Device (LHD), and a position-dependent variation in the directionality of deposition layer formation was found. (author)

[en] An experimental study of the local island divertor (LID) was done in the Compact Helical System (CHS). The separatrix of an m/n = 1/1 magnetic island, formed at the edge region by additional magnetic field was utilized as a divertor configuration. The main purpose of this configuration is to control the neutral particle recycling. The reduced line averaged density and radiation power loss, and improved energy confinement were obtained under an LID configuration

[en] Since how dust is generated and how it is transported in nuclear fusion can affect the nuclear burning plasma performance and its retention, this has become an important research subject. How much of hydrogen isotopes can be contained in dust has also become an important research subject, from the viewpoint of tritium economy and safety in face of nuclear fusion reactor operation. This paper takes up the theme of 'nuclear burning plasma operation,' with a focus on not only the retention of nuclear burning plasma, but also the operation of nuclear fusion reactor. It explains these research projects, with a focus on the understanding that has been obtained so far and the current research situation. Attention is especially paid to metal dust, since the divertor plate of ITER has become whole tungsten, and the strong candidate of plasma facing material of nuclear fusion reactors is tungsten. In addition, this paper takes up the control and monitoring of dust accumulation amount in the vacuum vessel involved in the operation of a nuclear fusion reactor, and explains the study situation at ITER, which will become the closest to the nuclear burning plasma equipment. (A.O.)

[en] Fully three-dimensional simulation of neutral particle transport is applied to analyses of the density profiles of neutral hydrogen in the plasma periphery and in the divertor region on the large helical device (LHD). For considering the three-dimensionally complicated geometry of the LHD plasma and the vacuum vessel, a detailed three-dimensional grid model is formed. In order to simulate the effect of the full toroidal geometry of the plasma, a new wall 'warp wall' is introduced. The calculations of three-dimensional profile of neutral densities in the grid model prove the availability of the warp walls for detailed analyses of neutral transport. The simulation strongly suggests that baffle plates installed in the inner region located between two helical coils in the vacuum vessel are cost-effective in achieving divertor detachment and efficient particle removal and control in the plasma periphery. (author)

[en] The number density of collision points of energetic tritons and a plasma facing wall of the Large Helical Device (LHD) of the National Institute for Fusion Science is visualized by OpenGL triangles colored according to the number density in the virtual-reality (VR) space. The triangles form the divertor plates and vacuum vessel wall in the LHD, and the energetic tritons attack the plates and wall. The number of tritons interfering with them is counted in each triangle. The colored OpenGL triangles are superimposed onto the vacuum vessel wall rendered by Unity and they are displayed in one VR world in real time. We can research the region where the tritons attack on the wall and discuss the number density. The tritons attack the divertor plates and a part of the wall. The number density of almost the triangles is from 1 (1/mm²) to 5 (1/mm²). (author)

[en] In order to observe edge neutral recycling for fusion machine operation, the absolute values of H_α and HeI emissions have been usually analyzed with measurement of edge density and temperature. On the other hand, it is known that the temperature and density dependences of each line can be negated by taking the ratio between the two lines. The intensity ratio of H_α (6563 A) to HeI (5876 A or 6678 A) visible spectral lines is adopted, and the results are presented here instead of the method usually used for monitoring the edge particle recycling and the effect of wall conditioning in LHD. The ionization events per photon were calculated for both emissions using a collisional-radiative model, and the ratio of H⁺ flux (≡ H⁰ influx) to He⁺ flux (≡ He⁰ influx) was obtained. The H_α and HeI (5876 A) emissions in LHD have been measured using a monitor assembly with an interference filter and optical fiber array. The H⁺/He⁺ flux ratio has been thus evaluated by integrating the H_α and HeI emissions from a 10-channel toroidal array. As an example of the measurement, all discharges performed during the past 9 years in LHD have been analyzed. The present method shows that the after effects of the H₂ and He glow discharge cleaning for vacuum wall conditioning on the LHD discharges become clearly visible. The replacement of deposited atoms on the carbon divertor plates was also analyzed. It is found that the replacement is completed by 40-50 shot repetition of discharges. These results indicate the effectiveness of the present method as a good monitor for fusion machine operation. (author)