Recent Fusion Research in the National Institute for Fusion Science

The National Institute for Fusion Science (NIFS), which was established in 1989, promotes academic approaches toward the exploration of fusion science for steady-state helical reactor and realizes the establishment of a comprehensive understanding of toroidal plasmas as an inter-university research organization and a key center of worldwide fusion research. The Large Helical Device (LHD) Project, the Numerical Simulation Science Project, and the Fusion Engineering Project are organized for early realization of net current free fusion reactor, and their recent activities are described in this paper. The LHD has been producing high-performance plasmas comparable to those of large tokamaks, and several new findings with regard to plasma physics have been obtained. The numerical simulation science project contributes understanding and systemization of the physical mechanisms of plasma confinement in fusion plasmas and explores complexity science of a plasma for realization of the numerical test reactor. In the fusion engineering project, the design of the helical fusion reactor has progressed based on the development of superconducting coils, the blanket, fusion materials and tritium handling

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

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

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)

Three-dimensional neutral particle transport simulation for analyzing polarization resolved H-alpha spectra in the large helical device

Change of Hα intensity profiles depending on magnetic configurations is observed in the divertor plasma. It can be explained by the magnetic field line structures in the ergodic layer and the divertor legs. The behavior of neutral particles in the plasma periphery is investigated by a three-dimensional neutral particle transport simulation code which assumes that the distribution of the plasma flow onto the divertor plates corresponds to that of the strike points calculated by magnetic field line traces. Vertical Hα intensity profiles and polarization resolved Hα spectra are calculated by the simulation code including the effect of Doppler broadening, fine structure splitting and polarization of the Hα emission, which agree well with the measurements in various magnetic configurations. It shows spontaneous formation of high neutral density in inboard side of the torus, which is independent of the magnetic configurations in LHD

The Effect of Non-Axisymmetry of Magnetic Configurations on Radial Electric Field Transition Properties in the LHD

Transition property of the radial electric field (Er) in LHD have been theoretically investigated and also applied to explain experimental results. Especially, effects of the helicity of the magnetic configuration on the condition to realize the electron root are examined. Larger helicity makes the threshold collisionality higher. This is attributed to the nonlinear dependence of Γe(Er) in a low collisional regime. This interesting feature predicts that the threshold temperature becomes higher for a case of smaller helicity. The variation of the threshold density anticipated from the analysis for cases with different magnetic axis position is qualitatively verified in the density scan experiment

Investigation of the Helical Divertor Function and the Future Plan of a Closed Divertor for Efficient Particle Control in the LHD Plasma Periphery

The function of the divertor plasmas on the particle control in the plasma periphery is investigated from viewpoints of magnetic field line structures and neutral particle transport in the Large Helical Device (LHD). It shows that the particle and heat deposition on the divertor plate arrays are qualitatively explained by the distribution of strike points calculated by magnetic field line tracing including a particle diffusion effect. Control of neutral particle fueling from the divertor plates is a critical issue for sustaining long-pulse discharges and achieving superdense core plasmas. The behavior of neutral particles in the plasma periphery has been investigated by Hα emission measurements and a neutral particle transport simulation. It reveals that gas fueling from the toroidally distributed divertor plates heated by protons accelerated by ion cyclotron resonance frequency wave is necessary for explaining measurements in a long-pulse discharge, and the spatial profile of the neutral particle density in the plasma periphery in various magnetic configurations is explained by the strike point distribution. Based on these analyses, a closed helical divertor configuration optimized for the intrinsic magnetic field line structure in the plasma periphery is proposed for efficient particle control and heat load reduction on the divertor plates

Stability and Confinement Studies of High-Performance NBI Plasmas in the Large Helical Device Toward a Steady-State Helical Fusion Reactor

Recent progress in plasma performance and the understanding of the related physics in the Large Helical Device is overviewed. The volume-averaged beta value is increased with an increase in the neutral beam injection (NBI) heating power, and it reached 5.0% of the reactor-relevant value. In high-β plasmas, the plasma aspect ratio should be controlled so that the Shafranov shift would be reduced, mainly to suppress transport degradation and the deterioration of the NBI heating efficiency. The operational regime of a high-density plasma with an internal diffusion barrier (IDB) has been extended, and the IDB, which was originally found using the local island divertor, has been realized in the helical divertor configuration. The central density was recorded as high as 1 × 1021 m-3, and the central pressure reached 130 kPa. Based on these high-density plasmas with the IDB, a new ignition scenario has been proposed. This should be a scenario specific to the helical fusion reactor, in which the helical ripple transport would be mitigated. A low-energy positive-NBI system was newly installed for an increase in the direct ion heating power. As a result, the ion temperature (Ti) exceeded 5.2 keV at a density of 1.2 × 1019 m-3 in a hydrogen plasma. Transport analysis shows improvement of ion transport, and the Ti-increase tends to be accompanied by a large toroidal rotation velocity of the order of 50 km/s in the core region. The plasma properties in the extended operational regime are discussed from the perspective of a steady-state helical fusion reactor

Improvement of Ion Confinement in Core Electron-Root Confinement (CERC) Plasmas in Large Helical Device

An increase in ion temperature has been observed with superposition of centrally focused electron cyclotron resonance heating (ECRH) to plasmas heated by high-energy neutral beam injection (NBI) in Large Helical Device. The ion-temperature (Ti) rise is accompanied by the formation of electron internal transport barrier (ITB). A transport analysis shows that ion transport as well as electron transport is improved with the reduction of anomalous transport. A neoclassical ambipolar flux calculation shows a positive radial-electric field (Er) in the region of the Ti rise, and Er should suppress the enhancement of ripple transport due to the Ti-rise. These analyses indicate the ion transport improvement in the core electron-root confinement plasmas. Toroidal rotation is driven in the co-direction by applying ECRH, and the toroidal rotation velocity is increased with the Ti rise. A correlation between the Ti rise and toroidal rotation is suggested

増強されたECHアンテナシステムを用いたLHDにおけるECCD適用性の向上

The power injection system for electron cyclotron heating (ECH) and electron cyclotron current drive (ECCD) was modified and upgraded. An outside horizontal port 2-O on the Large Helical Device (LHD) was furnished with two antenna systems for the EC-waves of the frequencies of 77 and 154 GHz, respectively. In addition to them, two new antenna systems for 77 and 154 GHz waves were installed in the 2-O port. Each antenna in the 2-O port has wide range of EC-wave beam direction control so that these are suitable for ECCD which requires toroidally oblique EC-wave beam injection. In the LHD 18th experimental campaign in 2014-2015, an ECCD experiment with second harmonic resonance condition, on-axis magnetic field of 1.375 T for 77 GHz waves, was performed in which some combination patterns of two 77 GHz ECCDs were applied. The discharges of dual co- and dual counter-ECCDs showed remarkable plasma currents of ∼±26 kA in both of the co- and counter-directions, by 6 s pulse duration and injection powers of 366 and 365 kW. The new antenna has nearly the same capability for ECCD with that of the existing antenna. The improvement in the flexibility of the ways of applying plural ECCDs will offer a highly useful tool for investigations on the phenomena concerning with the plasma current such as magnetohydro-dynamics