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Optimization of the helically-winding structure required for the helical fusion reactor utilizing the high temperaturesuperconductor (HTS) tape is studied by numerical calculation. When the HTS tapes are wound to bethe helical structure without the torsion, a significant edgewise strain appears. This might lead to the degradationof the performance of the HTS tape. Therefore, the edgewise strain must be minimized. It is found that the edgewise strain can be suppressed by the proper determination of the torsion by the optimization utilizing the Fourier decomposition of the distribution of the edgewise strain. The HTS tapes with edgewise strain are intrinsically deformed due to the torsion to reduce the edgewise strain so to be an “edgewise-strain-free” helical shape.

A High-Temperature Superconducting (HTS) conductor and magnet impregnated with low-melting point metal named “WISE” (Wound and Impregnated Stacked Elastic tapes) has been invented. This paper reports results on the first trial of the WISE conductor/magnet and its testing. An HTS solenoid coil without turn-to-turn insulation in accordance with the WISE concept was fabricated and tested. The flexibility of HTS-WISE conductor allowed handmade coil winding. In the first charging test after impregnation, there was no degradation due to thermal contraction. However, it was observed that thermal cycle damaged the HTS tapes after the second cooling. Even with some degradation, the sample achieved a central magnetic field of 0.16 T, having an 800 A current in the steady-state. Despite the appearance of an electric field of ∼1 mV/m along the conductor, quench did not occur. In the overcurrent test, self-protection characteristics of a non-insulation coil appeared. These results show that WISE is a smart winding method for making robust HTS coils having even complex 3D shapes.

Edge MHD instabilities with pressure collapse are found in relatively high beta and low magnetic Reynolds number regions with a magnetic axis torus outward-shifted configuration of the large helical device (LHD), and characteristics and onset conditions of the instability are investigated. The instability has a radial structure with an odd parity around the resonant surface, which is different from that of the interchange instability typically observed in the LHD. The onset condition dependence on the magnetic axis location shows that the onset beta increases as the magnetic axis location moves more torus inwardly, and the instability appears only in limited configurations where the magnetic axis is located between 3.65 and 3.775 m. In such configurations, the resonant surface location is close to an index of the plasma boundary. This fact suggests that the distance between the resonant surface location and the plasma boundary plays an important role in the onset, and a possibility that the instability is driven by an external mode.

A High-Temperature Superconducting (HTS) magnet is being considered to for use in the next-generation helical experimental devices. Three types of large-current HTS conductors are being developed, and one of them is the STARS (Stacked Tapes Assembled in Rigid Structure) conductor which uses HTS tapes with a simple stacking technique. Following the proof-of-principle experimental results obtained in the former 100-kA-class prototype hand-made conductor sample, an actually applicable conductor is being developed with a rated current of 18 kA at a temperature of 20 K and a magnetic field of ∼10 T. One of the crucial requirements for this conductor is to have a high current density of 80 A/mm2. In the first phase of the development, a 3-m short sample was fabricated by applying laser-beam welding to the stainless-steel jacket. It was tested in liquid nitrogen at 77 K with no external magnetic field. Then the sample was tested in gaseous helium at 20 - 40 K under a magnetic field of 6 - 8 T, and the results show that the basic requirements were satisfied.

High-temperature superconducting (HTS) conductor is a feasible candidate to make magnets for the next generation fusion devices because of its higher temperature margins and higher critical current in a high magnetic field in comparison to low-temperature superconducting (LTS) conductors. The recently proposed concept of the HTS-WISE (Wound and Impregnated Stacked Elastic tapes) conductor was studied to clarify its characteristics under certain magnetic fields. The WISE conductor, including 30-stacked REBCO (Rare-Earth Barium Copper Oxide) tapes, was fabricated and energized in a 9-T test facility which produced the condition of magnetic field B = 5 - 8 T and a temperature T = 30 - 50 K. Obtained critical currents (5.4 - 10.8 kA) increased with a decreasing magnetic field and/or temperature under the condition of T > 40 K. The maximum current of 16.9 kA was obtained at T = 30 K, which corresponded to the engineering current density jE = 60 A/mm2. Experimental results showed qualitative agreement with numerical calculations of the critical current. We confirmed the operation of the WISE conductor under a high magnetic field and low temperature.

The core plasma confinement during detachment phase is investigated in the discharges with application of resonant magnetic perturbation (RMP) field in LHD. The RMP application creates a remnant magnetic island in the edge stochastic layer, which largely changes the plasma parameter profiles including impurity radiation. The electron temperature and pressure profiles are flattened at the island, while the electron density is slightly peaked at the edge of the island. The estimated impurity radiation profile is enhanced and fixed around the magnetic island during the detached phase, where the discharge is stably sustained with controlled level of radiation. Without RMP, the radiation penetrates the confinement region, leading to radiation collapse. It is found that in the case of the RMP application the plasma stored energy increases discontinuously at the detachment transition. In spite of the reduced effective plasma volume caused by the edge magnetic island and by the enhanced radiation there, the central plasma pressure finally exceeds the case without RMP. This is caused by the pressure profile peaking at the central region in the case with RMP. These results indicate clear change of core plasma confinement during the detached phase with RMP.

The internal mode structure of a precursor with the tearing-parity structure of the locked-mode-like instability is investigated. For the first time, the Thomson scattering system with high temporal and spatial resolution enables us to find the non-rotating temperature flattening region in the torus outboard side in addition to the rotating island during the slowing-down phase. Additionally, the width of the flattening region is 10% normalized by the plasma minor radius. Furthermore, the radial profiles of the pressure degradation during the slowing-down phase are evaluated. At the beginning of the slowing-down phase, the pressure degradation area is located slightly inside the resonant surface. After that, the peak location moves to the core region. Finally, the degradation area peaks at the plasma center.

The slowing-down mechanism of the locked-mode-like instabilities with and without an island structure is investigated through the effects of an external RMP (resonant magnetic perturbation) on the instabilities. For both instabilities, the slowing-down duration decreases with the increase in the external RMP, and the RMP dependence is consistent with the braking model of the j × B force due to the interaction between the instabilities and the external RMP. Moreover, the relationship between the amplitude and the frequency of both locked-mode-like instabilities during the slowing down is consistent with the force balance model between the j × B force due to the external RMP and a viscous force. These results suggest that the slowing down of both locked-mode-like instabilities with finite external RMP occurs due to the j × B force driven by the external RMP.

In the Force Free Helical Reactor FFHR-d1 (R∼15.7 m, a∼2.5 m, Bo ∼4.5 T, 〈β〉∼5 % and the fusion power of 3 GW) [1, 2], it is demonstrated that the thermally unstable operation is better to achieve the higher density and lower temperature plasma (n(0)∼ 9 × 1020 m−3 and T(0)∼7 keV) than the impurity injection method. One drawback of the thermally unstable operation is a possibility of the thermal runaway when fueling systems have failures. However, we have found that the inherently safe function exists owing to the plasma outward shift during the thermal runaway. Thus we continue to pursue this high-density operation scenario for FFHR-d1. Preliminary estimation of the alpha energy loss fraction of ∼3 % is obtained for a broad density profile in the high-density ignition regime.

The deuterium (D) operation was initiated in the LHD in 2017. In the first campaign of the D experiments, we successfully extended the high temperature regime in the LHD. The new record of the ion temperature (Ti) of 10 keV associated with the ion internal transport barrier (ITB) was achieved due to several operational optimization. The thermal confinement characteristics of ITB plasmas were compared between hydrogen and D discharges. The effective ion thermal diffusivity of the ion-ITB plasmas was found to be smaller in the D discharges than that in the H discharges. The profiles of the Ti, the electron density, and the impurity of the high Ti plasmas strongly depended on the magnetic configuration and these profiles tended to peaked in the inward-shifted configuration. It was also found that the electron thermal confinement of the electron-ITB plasmas was clearly improved in the deuterium case. The GKV simulation showed the linear growth rate of TEM/ITG reduced in the plasmas with D both for the ion ITB and the electron ITB plasmas and qualitatively agreed with the tendency of the change in the thermal diffusivity obtained from the power balance analysis.

In magnetically confined experiments the presence of rational surfaces modifies the magnetic configuration with one or more islands embedded in the main plasma. The islands strongly modify particle and heat fluxes, hence their role must be included in the study of the transport properties. Two examples of this phenomenology are here presented: a cold pulse propagation in LHD with an RMP (1,1) island highlights a diffusivity reduction inside the island. A second study has been carried out on the RFX-mod reversed field pinch device where the transport on the ITB associated to a spontaneous m = 1 magnetic structure is discussed. To perform these studies a new approach in modeling transport in the presence of islands is presented: this new method, dubbed the multiple domains scheme (MDS), with some simplifying assumptions, is directly applicable to 1.5D transport codes. In the MDS different regions are identified inside the plasma all interfaced by the separatrix that acts as common boundary. In each domain a monotonic radial coordinate can be chosen and the metric elements can be computed accordingly. This scheme is independent of the equilibrium magnetic configuration hence it can be applied to either tokamaks, stellarators or reversed field pinches. The MDS has been implemented in a new numerical tool, dubbed the multiple axis solver.

We investigate the penetration threshold of resonant magnetic perturbation (RMP) by the external coils in the Large Helical Device (LHD) for various plasma aspect ratio configurations. The qualitative dependence on the collisionality is opposite to that in a high plasma aspect configuration; this is a quite unique property first found in the LHD. We also investigate the threshold dependence on the ion species, and find that the threshold in deuterium discharges is much smaller than that in hydrogen discharges. In all cases the thresholds are higher as the poloidal rotation becomes faster, which shows that poloidal rotation is the dominant driver to the RMP shielding. This is qualitatively consistent with the torque balance model between the electromagnetic and poloidal viscous torques. In a configuration of the LHD, the dependence of the threshold on the density is qualitatively similar to that in Ohmic tokamak plasmas, but the beta dependence is opposite to that of tokamaks. The difference arises from the cause of the viscous torque.

Spatially-resolved spectral measurement of a hydrogen pellet ablation cloud has been conducted in the Large Helical Device. Least square fittings of one of the obtained spectrum are inconsistent when the complete local thermodynamic equilibrium (CLTE) condition is supposed. A better agreement is found when a more general spectral model is used. The electron density of 1.4 × 1023 m−3 obtained is similar to that obtained in a previous study [M. Goto et al., Plasma Phys. Control. Fusion 49, 1163 (2007)]. The electron temperature of 0.4 eV is however significantly lower than the value of 1 eV derived in the same previous study.

The mode slowing down mechanism of the locked-mode-like instability without a large magnetic island is investigated, based on the LHD experimental analysis. The mode frequency coincides with E × B rotation frequency at the resonant surface and the slowing down is caused by two processes. One is the resonant surface moving to the small E × B rotation frequency region and the other is the slowing down of the E × B rotation frequency near the resonant surface. Both processes are almost the same as those of the locked-mode-like instability with a large magnetic island. The results suggest that the slowing down occurs even though the precursor does not have a large magnetic island. However, when the external RMP is imposed, the mode frequency in the slowing down phase sometimes does not coincide with the E × B rotation frequency. Moreover, the mode amplitude during the slowing down phase increases with the decrease of the mode frequency both with and without the imposed external RMP, which suggests that the instability growth in the slowing down phase is more strongly related to the mode frequency than the E × B rotation frequency because the mode sometimes does not rotate with the E × B rotation.

The accumulation and behavior of impurities is one of the most important subjects in the development of magnetically confined fusion reactors because impurities can potentially cause cooling and worsen the confinement of the hot core plasma. Tracer-encapsulated solid pellets (TESPELs) have demonstrated some results for impurity injection for fusion-reactor plasma studies [N. Tamura et al., J. Phys. Conf. Ser. 823, 012003 (2017)]. However, the TESPEL technique has several shortcomings, for example, the penetration depth and the amounts of tracer impurities. In the present study, we have developed a tracer-containing, compact-toroid (TCCT) injection system that utilizes a magnetized coaxial plasma gun (MCPG). The discharge current through the MCPG sputters and ionizes the electrode material, and the Lorenz self-force accelerates it as a plasmoid. The MCPG easily accelerates a magnetized plasmoid to speeds greater than the ion thermal velocity of several tens of kilometers per second. The accelerated and ejected plasmoid that contains the tracer ions is itself a warm, ionized plasma. Therefore, a TCCT can potentially be injected into the core region of a target plasma with less adverse effect.

Spontaneous dynamics of magnetic islands depending on various plasma parameters in the Large Helical Device (LHD) are described. The structure of the magnetic island undergoes deformation during a ...

A spectroscopic method for spatial resolution measurement in fuel pellet ablation clouds is being developed in the Large Helical Device (LHD). Spatial resolution is obtained thanks to optics that have a narrow, band-shaped field-of-view. The Stark-broadened H&beta, emission line of a deuterium pellet ablation cloud is isolated and analyzed with a spectral lineshape code. The electron density profile of the ablation cloud along its direction of elongation is derived through least squares fitting. The obtained profile is peaked and has a dip at its center which confirms what can be found in simulations. Moreover, the order of magnitudes for the derived electron densities are in agreement with what has already been found in the LHD.

Local transport properties of high-beta plasmas in the Large Helical Device are studied by comparing the beta dependence of the experimental results with that of the gyro-reduced Bohm-type transpor...

The interaction between pressure driven modes and magnetic islands generated by a resonant magnetic perturbation (RMP) in the large helical device (LHD) is numerically analyzed. In this analysis, three-dimensional treatment is essential in the equilibrium and dynamics calculations, because the equilibrium pressure profile is deformed by the RMP. The deformation changes the linear mode structure from the interchange type to the ballooning-like type that is localized around the X-point of the island in the equilibrium magnetic field including the RMP. This mode causes a pressure collapse in the nonlinear evolution, which spreads from the X-point to the core. Therefore, the spatial phase of the collapse is fixed to the island geometry. The fixed phase agrees with the LHD experimental results with a natural error field.

This paper describes recent progress in magnetohydrodynamics (MHD) study of high-beta plasmas of the Large Helical Device. Control of the plasma aspect ratio (Ap) in the range of 6.3 to 8.3 was done in order to optimize the configuration for high-beta plasma production and to investigate the MHD characteristics. The experiments brought a maximum average beta of 4.3% at the Ap = 6.6 configuration. MHD activities in the periphery are dominantly observed in such a high-beta region, and their amplitudes increase with decreasing magnetic Reynolds number (S) and have clear dependence on the S parameter. When the plasma aspect ratio is increased, minor collapse due to the m/n = 1/1 mode without rotation occurs. It is enhanced further by the plasma current reducing magnetic shear and degrades the beta value by >50%. The results are expected to give important information on the operation regime and the future design of the helical fusion reactor and to contribute to experimental knowledge of ideal and resistive instability.

One of the main objectives of the LHD is to extend the plasma confinement database for helical systems and to demonstrate such extended plasma confinement properties to be sustained in steady state. Among the various plasma parameter regimes, the study of confinement properties in the collisionless regime is of particular importance. Electron cyclotron resonance heating (ECRH) has been extensively used for these confinement studies of the LHD plasma from the initial operation. The system optimizations including the modification of the transmission and antenna system are performed with the special emphasis on the local heating properties. As the result, central electron temperature of more than 10 keV with the electron density of 0.6 x 10$^{19}$ m$^{-3}$ is achieved near the magnetic axis. The electron temperature profile is characterized by a steep gradient similar to those of an internal transport barrier observed in tokamaks and stellarators. 168 GHz ECRH system demonstrated efficient heating at over the density more than 1.0 x 10$^{20}$ m$^{-3}$. CW ECRH system is successfully operated to sustain 756 s discharge.

Fixed boundary three-dimensional anisotropic pressure equilibria are computed that are consistent with a family of distribution functions that load particles preferentially towards v parallel to similar or equal to v and model the effects of balanced tangential neutral beam injection. The total pressure surfaces remain coupled with the magnetic surfaces even when the parallel beam pressure greatly exceeds its perpendicular counterpart. The anisotropy factors and the Shafranov shift, which increases with the anisotropy, are calculated in a currentless two-field period quasiaxisymmetric stellarator device.

In the low-density plasmas of the Large Helical Device, the shape of the electron temperature profile changes depending on the direction of the tangential neutral beam injection (NBI) when the magnetic axis position is inward-shifted at R = 3.50 m. Core flattening was observed in plasmas heated by counter-NBI. The electron thermal diffusivities in co-NBI and counter-NBI-heated plasmas are compared. The diffusivity becomes large at the central region in the case of counter-NBI. This result shows that the flattening in the electron temperature profile is not caused simply by a change in the power deposition only. Some magnetic fluctuations are seen during counter-NBI. On the other hand, it is a promising feature that the electron thermal diffusivity at the peripheral region does not increase with the heating power in co-NBI plasmas.

Long-pulse operation and high-energy particle confinement properties were studied using ion cyclotron range of frequency (ICRF) heating for the Large Helical Device. For the minority-ion mode, ions with energies up to 500 keV were observed by concentrating the ICRF heating power near the plasma axis. The confinement of high-energy particles was studied using the power-modulation technique. This confirmed that the confinement of high-energy particles was better with the inward-shifted configuration than with the normal configuration. This behavior was the same for bulk plasma confinement. Long-pulse operation for more than 2 min was achieved during the experimental program in 2002. This was mainly due to better confinement of the helically trapped particles and accumulation of fewer impurities in the region of the plasma core, in conjunction with substantial hardware improvements. Currently, the plasma operation time is limited by an unexpected density rise due to outgassing from the chamber materials. The temperature of the local carbon plates of the divertor exceeded 400 deg, C, and a charge-coupled device camera observed the hot spots. The hot spot pattern was well explained by a calculation of the accelerated-particle orbits, and those accelerated particles came from outside the plasma near the ICRF antenna.

Characteristics of transport in electron internal transport barriers (ITB) and in the vicinity of a rational surface with a magnetic island are studied with transient transport analysis as well as with steady state transport analysis. Associated with the transition of the radial electric field from a small negative value (ion-root) to a large positive value (electron-root), an electron ITB appears in the Large Helical Device [M. Fujiwara et al., Nucl. Fusion 41, 1355 (2001)], when the heating power of the electron cyclotron heating exceeds a power threshold. Transport analysis shows that both the standard electron thermal diffusivity, chie, and the incremental electron thermal diffusivity, chieinc (the derivative of normalized heat flux to temperature gradient, equivalent to heat pulse chie), are reduced significantly (a factor 5 10) in the ITB. The chieinc is much lower than the chie by a factor of 3 just after the transition, while chieinc is comparable to or even higher than chie before the transition, which results in the improvement of electron transport with increasing power in the ITB, in contrast to its degradation outside the ITB. In other experiments without an ITB, a significant reduction (by one order of magnitude) of chieinc is observed at the O-point of the magnetic island produced near the plasma edge using error field coils. This observation gives significant insight into the mechanism of transport improvement near the rational surface and implies that the magnetic island serves as a poloidally asymmetric transport barrier. Therefore the radial heat flux near the rational surface is focused at the X-point region, and that may be the mechanism to induce an ITB near a rational surface.

We analyze the relationship between the experimentally observed pressure gradients at resonant rational surfaces and the theoretically predicted ideal magnetohydrodynamics (MHD) unstable region of global modes in the large helical device (LHD). According to the stability analysis of the ideal MHD modes with a low toroidal mode number, we find that the ideal MHD mode gives a constraint on the operational regime of the pressure gradients in the core. In the edge, a clear saturation of the pressure gradients due to the ideal MHD instability has not been observed up to the high beta regime around 3% as the volume-averaged toridal beta value, where global ideal MHD modes are predictedto be unstable.

he Large Helical Device (LHD) has been extending an operational regime of net-current free plasmas towardsthe fusion relevant condition with taking advantage of a net current-free heliotron concept and employing a superconducting coil system. Heating capability has exceeded 10 MW and the central ion and electron temperatureshave reached 7 and 10 keV, respectively. The maximum value of β and pulse length have been extended to 3.2% and 150 s, respectively. Many encouraging physical findings have been obtained. Topics from recent experiments, which should be emphasized from the aspect of theoretical approaches, are reviewed. Those are (1) Prominent features in the inward shifted configuration, i.e., mitigation of an ideal interchange mode in the configuration with magnetic hill, and confinement improvement due to suppression of both anomalous and neoclassical transport, (2) Demonstration ofbifurcation of radial electric field and associated formation of an internal transport barrier, and (3) Dynamics of magnetic islands and clarification of the role of separatrix.

Ion heating experiments have been carried out in the large helical device using ECH (82.5, 84.0, 168 GHz, ≤1 MW), ICRF (38.5 MHz, ≤2.7 MW) and NBI (H° beam: 160 keV, ≤9 MW). The central ion temperature has been observed from the Doppler broadening of Ti XXI (2.61 A) and Ar XVII (3.95 A) x-ray lines, which are measured using a newly installed crystal spectrometer with a charge-coupled device. Recently, in ECH discharges, on-axis heating became possible. As a result, a high Te(0) of 6–10 keV and a high ion temperature of 2.2 keV were obtained at ne = 0.6×1013 cm−3. A clear increment of Ti was also observed with the enhancement of the electron–ion energy flow when the ECH pulse was added to the NBI discharge. These results demonstrate the feasibility towards ECH ignition. A clear Ti increment was observed also in ICRF discharges at low density ranges of (0.4–0.6)×1013 cm−3 with appearance of a new operational range of Ti(0) = 2.8 keV > Te(0) = 1.9 keV. In low power ICRF heating (1 MW), the fraction of bulk ion heating is estimated to be 60% of the total ICRF input power, which means Pi>Pe. Higher Ti(0), up to 3.5 keV, was obtained for a combined heating of NBI ( Te(0), whereas the Ti(0) remained at relatively low values of 2 keV in H2 and He NBI discharges due to less Pi. The main reasons for the high Ti achievement in the Ne discharges are: (1) 30% increment of deposition power, (2) increase in Pi/ni (five times, Pi/niPe/ne, Pi

An internal transport barrier (ITB) was observed in the electron temperature profile in the Large Helical Device [O. Motojima et al., Phys. Plasmas 6, 1843 (1999)] with a centrally focused intense electron cyclotron resonance microwave heating. Inside the ITB the core electron transport was improved, and a high electron temperature, exceeding 10 keV in a low density, was achieved in a collisionless regime. The formation of the electron-ITB is correlated with the neoclassical electron root with a strong radial electric field determined by the neoclassical ambipolar flux. The direction of the tangentially injected beam-driven current has an influence on the electron-ITB formation. For the counter-injected target plasma, a steeper temperature gradient, than that for the co-injected one, was observed. As for the ion temperature, high-power NBI (neutral beam injection) heating of 9 MW has realized a central ion temperature of 5 keV with neon injection. By introducing neon gas, the NBI absorption power was increased in low-density plasmas and the direct ion heating power was much enhanced with a reduced number of ions, compared with hydrogen plasmas.

The superconducting machine LHD has conducted long pulse experiments for four years to achieve long-duration plasmas with high performance. The operational regime was largely extended in discharge duration and plasma density. In this paper, the plasma characteristics, in particular, plasma performance and impurity behaviour in long pulse discharges are described. Confinement studies show that global energy confinement times are comparable to those in short pulse discharges. Long sustainment of high performance plasma, which is equivalent to the previous achievement in other devices, was demonstrated. Long pulse discharges enabled us to investigate impurity behaviour in a long timescale. Intrinsic metallic impurity accumulation was observed in a narrow density window (2–3×1019 m−3) only for hydrogen discharges. Impurity transport study by using active impurity pellet injection shows a long impurity confinement time and an inward convection in the impurity accumulation window, which is consistent with the intrinsic impurity behaviour. The pulsed neon gas injection experiment shows that the neon penetration into the plasma core is caused by the inward convection due to radial electric field. Finally, impurity accumulation control with an externally induced magnetic island at the plasma edge was demonstrated.

As the finalization of a hydrogen experiment towards the deuterium phase, the exploration of the best performance of hydrogen plasma was intensively performed in the large helical device. High ion and electron temperatures, Ti and Te, of more than 6 keV were simultaneously achieved by superimposing high-power electron cyclotron resonance heating onneutral beam injection (NBI) heated plasma. Although flattening of the ion temperature profile in the core region was observed during the discharges, one could avoid degradation by increasing the electron density. Another key parameter to present plasma performance is an averaged beta value $\left\langle \beta \right\rangle $ . The high $\left\langle \beta \right\rangle $ regime around 4% was extended to an order of magnitude lower than the earlier collisional regime. Impurity behaviour in hydrogen discharges with NBI heating was also classified with a wide range of edge plasma parameters. The existence of a no impurity accumulation regime, where the high performance plasma is maintained with high power heating >10 MW, was identified. Wide parameter scan experiments suggest that the toroidal rotation and the turbulence are the candidates for expelling impurities from the core region.

This paper reports the main recent results of the RFX-mod fusion science activity. The RFX-mod device is characterized by a unique flexibility in terms of accessible magnetic configurations. Axisymmetric and helically shaped reversed-field pinch equilibria have been studied, along with tokamak plasmas in a wide range of q(a) regimes (spanning from 4 down to 1.2 values). The full range of magnetic configurations in between the two, the so-called ultra-low q ones, has been explored, with the aim of studying specific physical issues common to all equilibria, such as, for example, the density limit phenomenon. The powerful RFX-mod feedback control system has been exploited for MHD control, which allowed us to extend the range of experimental parameters, as well as to induce specific magnetic perturbations for the study of 3D effects. In particular, transport, edge and isotope effects in 3D equilibria have been investigated, along with runaway mitigations through induced magnetic perturbations. The first transitions to an improved confinement scenario in circular and D-shaped tokamak plasmas have been obtained thanks to an active modification of the edge electric field through a polarized electrode. The experiments are supported by intense modeling with 3D MHD, gyrokinetic, guiding center and transport codes. Proposed modifications to the RFX-mod device, which will enable further contributions to the solution of key issues in the roadmap to ITER and DEMO, are also briefly presented. © 2017 IAEA, Vienna.

In order to clarify the mechanism responsible for slowing down the precursor of an instability in Large Helical Device (LHD), whose behavior is similar to the locked mode instability in tokamaks, the spatial structure of the precursor of the locked-mode-like instability, and the relationship between the rotation of the precursor and the E × B rotation were experimentally investigated. The precursor rotates together with the E × B rotation at the resonant surface, and the precursor rotation slows down because of a decrease of the E × B rotation. The multi-channel fluctuation measurement of the precursor suggests that the precursor has a magnetic island, which may be related to the decrease of the E × B rotation. In addition, the reason for the appearance of the precursor with a magnetic island is discussed. The precursor appears when a magnetic island grows initially without rotation but then shrinks and begins to rotate.

Since the start of the Large Helical Device (LHD) experiment, various attempts have been made to achieve improved plasma performance in LHD [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)]. Recently, an inward-shifted configuration with a magnetic axis position R_ax of 3.6 m has been found to exhibit much better plasma performance than the standard configuration with R_ax of 3.75 m. A factor of 1.6 enhancement of energy confinement time was achieved over the International Stellarator Scaling 95. This configuration has been predicted to have unfavorable magnetohydrodynamic (MHD) properties, based on linear theory, even though it has significantly better particle-orbit properties, and hence lower neoclassical transport loss. However, no serious confinement degradation due to the MHD activities was observed, resolving favorably the potential conflict between stability and confinement at least up to the realized volume-averaged beta of 2.4%. An improved radial profile of electron temperature was also achieved in the configuration with magnetic islands, minimized by an external perturbation coil system for the Local Island Divertor (LID). The LID has been proposed for remarkable improvement of plasma confinement like the high (H) mode in tokamaks, and the LID function was suggested in limiter experiments.

The plasma density of a field reversed configuration (FRC) needs to be decreased below the present experimental regime in order to heat the plasma and sustain the configuration by a high energy neutral beam in an FRC reactor. However, as the plasma is produced in a linear vacuum vessel, there exists a severe breakdown limit at a low fill pressure as compared with a toroidal system. A method to form FRCs beyond the breakdown limit is proposed here. The preionized plasma is compressed by a strong bias field to enhance the plasma flow from the confinement region to the outside region and is then diluted before the start of the confinement field on the NUCTE device. The use of a diluted plasma enables the critical density of the FRC to be lowered from 1.1 × 1021 to 5.6 × 1020m-3 and the sum of the electron and ion temperatures to be increased from 0.35 to 0.81 keV.

Bifurcation physics of a magnetic island was investigated using the heat pulse propagation technique produced by the modulation of electron cyclotron heating. There are two types of bifurcation phenomena observed in a large helical device (LHD) and DIII-D. One is a bifurcation of the magnetic topology between nested and stochastic fields. The nested state is characterized by the bi-directional (inward and outward) propagation of the heat pulse with slow propagation speed. The stochastic state is characterized by the fast propagation of the heat pulse with electron temperature flattening. The other bifurcation is between the magnetic island with larger thermal diffusivity and that with smaller thermal diffusivity. The damping of toroidal flow is observed at the O-point of the magnetic island both in helical plasmas and in tokamak plasmas during a mode locking phase with strong flow shears at the boundary of the magnetic island. Associated with the stochastization of the magnetic field, the abrupt damping of toroidal flow is observed in LHD. The toroidal flow shear shows a linear decay, while the ion temperature gradient shows an exponential decay. This observation suggests that this flow damping is due to the change in the non-diffusive term of momentum transport.

Effects of poloidal shear flow on the stability of interchange modes in a Large Helical Device (LHD) configuration are numerically studied. Three-dimensional (3D) numerical codes are utilized for the equilibrium and stability calculations. A static equilibrium is employed and a model poloidal flow as a flux function is incorporated in the initial perturbation. The results show that the initially applied flow can suppress the growth of the interchange mode if the flow is sufficiently large.

This paper presents recent experimental results in the Large Helical Device (LHD), which is the world largest helical system (Rax~3.6 m, aav~0.6 m, Bax~2.75 T) and has capability of steady state operation because of the superconducting magnets. LHD confines the plasma with Tio~2 keV and ne ~0.8times1019 m-3 for 1905 sec using the ICRF heating. In this case, the total plasma heating energy was ~1.3 GJ. This indicates that high energy ions generated by ICRF is well confined. Large magnetic island in the vacuum field due to the error field is vanished in plasma with higher temperature or with higher beta, so the error field is not an issue in helical systems. The obtained averaged beta is more than 4.3 % in LHD. Since the beta is not limited by the MHD events or magnetic island but the beta is limited by the lack of heating power, further improvement of beta can be expected. Internal and external transport barrier have been also observed, and the electron temperature and the ion temperature exceed 10 keV in the improved confinement mode in LHD. Besides the potential of the disruption free steady stated operation, LHD experiment accumulates evidences so that fusion reactor becomes more realistic

Sawtooth oscillations have been observed in current-carrying helical plasmas by using electron-cyclotron-emission diagnostics in the Large Helical Device. The plasma current, which is driven by neutral beam injection, reduces the beta threshold of the sawtooth oscillation. When the central q value is increased due to the plasma current, the core region crashes, and, when it is decreased, the edge region crashes annularly. Observed rapid mixture of the plasma in the limited region suggests that these sawtooth crashes are reconnection phenomena. Unlike previous experiments, no precursor oscillation has been observed.

The hysteretic magnetic island response to an externally applied resonant magnetic perturbation (RMP) field is observed in the LHD. Thresholds of the amplitude of the RMP for the growth/healing transition of the magnetic island differ. In the case that the RMP is ramped up, that field is initially shielded and the magnetic island is healed. After that, when the increasing RMP exceeds a threshold, that field penetrates into the plasma and the magnetic island appears. In the case that the RMP is ramped down, the threshold of the RMP for island healing is smaller than that for growth.

Property of pressure driven modes in Large Helical Device (LHD) plasmas with a resonant magnetic perturbation (RMP) is numerically studied. Particularly, we analyze three-dimensional (3D) RMP effects on the linear magnetohydrodynamic (MHD) stability of the modes. For this purpose, 3D numerical codes are utilized for both calculations of an equilibrium including an RMP generating an m = 1/n = 1 magnetic island and the stability of the perturbations resonant at the ι = 1 surface. Here, m and n are the poloidal and the toroidal mode numbers, respectively, and ι denotes rotational transform. Owing to the RMP, the pressure driven mode is localized around the X-point of the island. The type of the mode structure changes from the interchange type to the ballooning type. This property is attributed to the fact that the equilibrium pressure gradient is larger at the X-point than at the O-point.

In the Large Helical Device experiments, a significant pressure gradient is observed in the edge region where the three-dimensional magnetohydrodynamics equilibrium analysis predicts the stochastization of magnetic field lines. Therefore, experimental investigation of this edge plasma behavior is attracting much interest to identify either the plasma response to stochastic magnetic fields or the topological change of magnetic fields due to the plasma. In this study, we have investigated the relation of the stochastization of magnetic field lines to the radial electric field, Er. When magnetic field lines become stochastic or open and connected to the vessel, electrons are lost along these field lines while ions are trapped. Then, a strong Er shear from negative to positive appears at the boundary. Two magnetic configurations with different widths of the stochastic layer in the vacuum are studied for low-? discharges. It has been found that the position of a strong Er shear appears outside of the last closed flux surface. When comparing the vacuum magnetic field, the positions of strong Er shear are observed in the edge of the stochastic layer. In the stochastic region, the scatter of stochastic field lines is studied. The position of a strong Er shear appears in the region where the outward scatter of the field lines increases. In that region, stochastic field lines intersect the regions with long and short connection lengths. Although the connection length is longer than the electron mean free path, the scatter of stochastic field lines with short connection lengths appear in the regions with positive Er. These results are compared with moderate-? discharges, where a large plasma response is expected. The prediction of the vacuum magnetic field qualitatively agree with the experimental observations but quantitative differences are found.