Observation of a Rotating Radiation Belt in LHD

A poloidally rotating radiation belt with helical structure was observed during the high density discharges with detachment by photodiode fan arrays and a fast camera in LHD. The peak of radiation rotates inside the last closed flux surface, and the direction and mode number of the poloidal rotation are electron diamagnetic and one, respectively. During the recombination phase after termination of the plasma heating, the rotation continues, and its rotating radius shrinks with shrinking of the plasma column. The poloidal rotating frequency depends on the heating power, and increases from the orders of several tens of Hz to several hundreds of Hz with shrinking of the rotation radius. The mechanism of the rotation remains uncertain

Study of carbon dust formation and their structure using inductively coupled plasmas under high atomic hydrogen irradiation

金沢大学理工研究域電子情報学系Experiments on erosion and dust formation on graphite materials have been performed using high power induction plasmas containing high atomic hydrogen flux (∼1024 m-2 s-1). Chemical sputtering by atomic hydrogen irradiation with incident energy below 1 eV eroded the graphite targets significantly, and the sputtering yield was roughly estimated to be 0.002-0.005, which is as high as that obtained by ion beam and fusion plasma experiments. The transport of the released hydrocarbon along the gas flow, interacting with low temperature plasmas, results in carbon dust formation on the eroded graphite target and also on the silicon and graphite samples located at the remote position. The dust size and density observed on the samples decreases with distance from the graphite target. The dust shape strongly depends on the target surface temperature, and the graphite dust turns into polyhedral particle like diamond when the surface temperature rises to 1100 K. © 2009 Elsevier B.V. All rights reserved

Removal of carbon deposited film and hydrogen retention control by low temperature H-C-N reactive plasmas

Control of tritium retention and its removal from the first wall of future fusion devices are the most crucial issues for safety and effective use of the fuel. Nitrogen injection into edge plasmas has been considered and tested as an effective method for suppression of carbon dust growth and reduction of hydrogen isotope inventory. In this paper we have investigated scavenger effects of nitrogen injected into H2/CH4 plasmas using a small helical device where low density (ne ∼ 1016 m-3) and low temperature (Te = 5-10 eV) hydrogen plasmas are generated in steady state condition like remote plasmas in fusion devices. It is shown from the comparison of the carbon film deposition and particle growth between those with and without N2 injection that the chemical erosion effects of nitrogen gas on the carbon film and particle growth strongly depends on the surface temperature. With increasing the surface temperature higher than ∼400 K, the nitrogen chemical erosion significantly works to reduce the hydrocarbon deposition. © 2013 Elsevier B.V. All rights reserved

Blob/Hole Generation in the Divertor Leg of the Large Helical Device

We have analyzed ion saturation current fluctuation measured by a fast scanning Langmuir probe (FSP) in edge region of the Large Helical Device (LHD). Positive and negative spikes of the ion saturation current were observed in the private region and on the divertor leg, respectively. It was found that the boundary position between these regions corresponds to the low-field side (LFS) edge of the divertor leg where the gradient of the ion saturation current profile was the maximum. Such a positional relationship resembles that near the separatrix in the LFS in tokamaks, where blobs and holes are generated. Statistical analysis indicates similar fluctuation characteristics among different magnetic devices

Validation of the plasma-wall interaction simulation code ERO2.0 by the analysis of tungsten migration in the open divertor region in the Large Helical Device

Tungsten migration in the open divertor region in the Large Helical Device is analyzed for validating the three-dimensional plasma-wall interaction simulation code ERO2.0. The ERO2.0 simulation reproduced the measurement of localized tungsten migration from a tungsten-coated divertor plate installed in the inboard side of the torus. The simulation also explained the measurement of the high tungsten areal density in the private side on a carbon divertor plate, next to the tungsten-coated divertor plate, by the tungsten prompt redeposition in plasma discharges for a low magnetic field strength in a counterclockwise toroidal direction. However, the simulation disagreed with the measurement of low tungsten areal density on the plasma-wetted areas on the carbon divertor plates, which indicated that the actual erosion rate of the redeposited tungsten should be much higher than that used in the ERO2.0 code

First EMC3-EIRENE Simulations with Divertor Legs of LHD in Realistic Device Geometry

An extended mesh system for EMC3-EIRENE has been developed to simulate peripheral plasma including the ergodic and the divertor leg regions of LHD. Both the open and the closed divertor configurations are available. A series of simulations for 8MW input power, five different electron densities at the LCFS (last closed flux surface) and the open/closed configurations were carried out. Approximately 10 times larger neutral pressure was observed under the dome structure compared with the open configuration, which is in good agreement with experimental measurements. In the case of the closed configuration, the leg regions have a large contribution of ionization to hydrogen recycling. In the case of high density discharges, however, electron temperature in the legs becomes low and the major contribution of ionization moves to the ergodic region. Significant influence of configurations is observed in the inboard side of LHD, where closed divertor components are installed but little influence is seen near the LCFS. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim

Effect of divertor legs on neutral particle and impurity retention for a closed helical divertor configuration in the Large Helical Device

A closed helical divertor (CHD) has been designed for efficient particle control in the plasma periphery and for retaining neutral particles and impurity ions in the divertor region. The effect of impurity retention by divertor legs for the CHD configuration is investigated from the viewpoints of neutral impurity transport and force balance of impurity ions along magnetic field lines. A fully three-dimensional neutral particle transport simulation proves that the plasma on the divertor legs is effective for retaining neutral particles/impurities in the CHD region. A one-dimensional impurity ion transport analysis predicts that friction force by plasma flow from the main plasma sweep impurity ions toward the divertor plates even in high neutral density case in which a steep temperature gradient is formed. It shows that the CHD configuration is promising for enhancing LHD plasma performance by effective control of the neutral particles and the impurity ions in the plasma periphery

Design of a Closed Helical Divertor in LHD and the Prospect for Helical Fusion Reactors

A new closed helical divertor configuration for efficient particle control and reduction of the heat load on the divertor plates is proposed. The closed divertor configuration practically utilizes an ergodic layer and magnetic field line configuration on divertor legs in helical systems. For optimization of the design of the closed divertor, the distribution of the strike points is calculated in various magnetic configurations in the Large Helical Device (LHD). It suggests that the installation of the closed divertor components in the inboard side of the torus under an inward shift configuration (Rax=3.60m) is the best choice for achieving the above two purposes. This divertor configuration does not interfere with plasma heating and diagnostic systems installed in outer ports. The prospect of the closed divertor configuration to a helical fusion reactor is investigated using a three-dimensional neutral particle transport simulation code with a one-dimensional plasma fluid calculation on the divertor legs. The investigation shows efficient particle pumping from the in board side and reduction of the heat load due to the combined effect of the optimized closed divertor geometry, ergodized divertor legs, and low electron temperature in the ergodic layer. It indicates a promising closed divertor configuration for helical fusion reactors

Study of the effect of a closed divertor configuration on neutral particle control in the LHD plasma periphery

An optimized closed divertor configuration for effective particle control in LHD is proposed from the viewpoints of the distribution of the strike points and neutral particle transport. Calculations of the distribution of the strike points indicate that 50% of the strike points locate in the inboard side of the torus in a standard magnetic configuration (Rax = 3.60 m). The ratio increases to 80% by installing target plates near lower/upper ports. A three-dimensional neutral particle transport simulation shows that installation of closed divertor components with the target plates raise the neutral pressure in the inboard side by more than one order of magnitude compared to that in the present open divertor case. The analysis of the neutral particle transport predicts that enhancement of the neutral pressure becomes moderate in outward shift configurations (Rax > 3.75 m)