Operation Of Large Helical Device Research Articles

Theoretical analysis of energetic-ion-driven resistive interchange mode stabilization strategies using a Landau closure model

The aim of the present study is to perform a theoretical analysis of different strategies to stabilize energetic-ion-driven resistive interchange mode (EIC) in Large Helical Device (LHD) plasma. We use a reduced MHD for the thermal plasma coupled with a gyrofluid model for the energetic particles (EP) species. The hellically trapped EP component is introduced through a modification of the drift frequency to include their precessional drift. The stabilization trends of the 1/1 EIC observed experimentally with respect to the thermal plasma density and temperature are reproduced by the simulations, showing a reasonable agreement with the data. The LHD operation scenarios with stable 1/1 EIC are identified, leading to the stabilization of the 1/1 EIC if the thermal plasma density and temperature are above a given threshold. The 1/1 EIC are also stabilized if the rotational transform is modified in a way that the 1/1 rational surface is located further away than 0.9 times the normalized radius, or the magnetic shear in the plasma periphery is enhanced. Also, LHD discharges with large magnetic fields show a higher EIC destabilization threshold with respect to the thermal plasma density. If the perpendicular NBI deposition region is moved further inward than 0.875 times the normalized radius the 1/1 EIC are also stabilized. In addition, increasing the perpendicular NBI voltage such that the EP energy is higher than 30 keV stabilizes the 1/1 EIC. Moreover, deuterium plasmas show a higher stability threshold for the 1/1 EIC than hydrogen plasmas. The experimental data shows a larger time interval between EIC events as the power of the tangential NBI is increased providing that the perpendicular NBI power is at least 13 MW. This implies a stabilizing effect of the tangential NBI.

Preparation and Commissioning for the LHD Deuterium Experiment

The deuterium experiment started from March 2017 on the large helical device (LHD) as a part of the LHD high-performance upgrade project. The objectives of the deuterium experiment are: 1) to realize the high-performance operation; 2) to explore the physics of isotope effect; 3) to demonstrate the confinement capability of energetic particles in helical devices; and 4) to explore the research on plasma–material interactions using the benefit of stable steady-state operation ability of LHD. As preparations for deuterium experiment, the positive-ion-based neutral beam injectors (NBIs) are upgraded to increase their injection power in deuterium operation. On the other hand, the injection power of negative-ion-based NBI is deteriorated due to the isotope effect of negative-ion source. The neutron diagnostic and the exhaust detritiation system are newly installed for the deuterium experiment. The commissioning of LHD for the deuterium experiment is quite successful. The high ion temperature operation region is extended by the deuterium experiment. Ion temperature exceeding 9 keV is achieved with the deuterium operation of LHD.

Investigation of irradiation effects on highly integrated leading-edge electronic components of diagnostics and control systems for LHD deuterium operation

High-temperature and high-density plasmas are achieved by means of real-time control, fast diagnostic, and high-power heating systems. Those systems are precisely controlled via highly integrated electronic components, but can be seriously affected by radiation damage. Therefore, the effects of irradiation on currently used electronic components should be investigated for the control and measurement of Large Helical Device (LHD) deuterium plasmas. For the precise estimation of the radiation field in the LHD torus hall, the MCNP6 code is used with the cross-section library ENDF B-VI. The geometry is modeled on the computer-aided design. The dose on silicon, which is a major ingredient of electronic components, over nine years of LHD deuterium operation shows that the gamma-ray contribution is dominant. Neutron irradiation tests were performed in the OKTAVIAN at Osaka University and the Fast Neutron Laboratory at Tohoku University. Gamma-ray irradiation tests were performed at the Nagoya University Cobalt-60 irradiation facility. We found that there are ethernet connection failures of programmable logic controller (PLC) modules due to neutron irradiation with a neutron flux of 3 × 106 cm−2 s−1. This neutron flux is equivalent to that expected at basement level in the LHD torus hall without a neutron shield. Most modules of the PLC are broken around a gamma-ray dose of 100 Gy. This is comparable with the dose in the LHD torus hall over nine years. If we consider the dose only, these components may survive more than nine years. For the safety of the LHD operation, the electronic components in the torus hall have been rearranged.

Feedback control of plasma density and heating power for steady state operation in LHD

For steady state operation, the feedback control of plasma density and heating power system was developed in the Large Helical Device (LHD). In order to achieve a record of the long pulse discharge, stable plasma density and heating power are needed. This system contains the radio frequency (RF) heating power control, interlocks, gas fueling, automatic RF phase control, ion cyclotron range of frequency (ICRF) antenna position control, and graphical user interface (GUI). Using the density control system, the electron density was controlled to the target density and using the RF heating power control system, the RF power injection could be stable. As a result of using this system, we achieved the long pulse up to 48min in the electron density of more than 1×1019m−3. Further, the ICRF hardware experienced no critical accidents during the 17th LHD experiment campaign in 2013.

The internal disruption as hard Magnetohydrodynamic limit of 1/2 sawtooth like activity in large helical device

Large helical device (LHD) inward-shifted configurations are unstable to resistive MHD pressure-gradient-driven modes. Sawtooth like activity was observed during LHD operation. The main drivers are the unstable modes 1/2 and 1/3 in the middle and inner plasma region which limit the plasma confinement efficiency of LHD advanced operation scenarios. The aim of the present research is to study the hard MHD limit of 1/2 sawtooth like activity, not observed yet in LHD operation, and to predict its effects on the device performance. Previous investigations pointed out this system relaxation can be an internal disruption [J. Varela et al., “Internal disruptions and sawtooth like activity in LHD,” 38th EPS Conference on Plasma Physics (2011), P5.077]. In the present work, we simulate an internal disruption; we study the equilibria properties before and after the disruptive process, its effects on the plasma confinement efficiency during each disruptive phase, the relation between the n/m = 1/2 hard MHD events and the soft MHD events, and how to avoid or reduce their adverse effects. The simulation conclusions point out that the large stochastic region in the middle plasma strongly deforms and tears the flux surfaces when the pressure gradient increases above the hard MHD limit. If the instability reaches the inner plasma, the iota profiles will be perturbed near the plasma core and three magnetic islands can appear near the magnetic axis. If the instability is strong enough to link the stochastic regions in the middle plasma (around the half minor radius ρ) and the plasma core (ρ<0.25), an internal disruption is driven.

Internal disruptions and sawtooth like activity in Large Helical Device

Large Helical Device (LHD) inward-shifted configurations are unstable to resistive magnetohydrodynamic (MHD) pressure-gradient-driven modes. These modes drive sawtooth like events during LHD operation. In this work, we simulate sawtooth like activity and internal disruptions in order to improve the understanding of these relaxation events and their effect over the device efficiency to confine the plasma, with the aim to improve the LHD present and future operation scenarios minimizing or avoiding the disadvantageous MHD soft and hard limits. By solving a set of reduced non-linear resistive MHD equations, we have studied the evolution of perturbations to equilibria obtained before and after a sawtooth like event in LHD. The equilibrium β value is gradually increased during the simulation until it reaches the experimental value. Sawtooth like events and internal disruption events take place in the simulation for β0 values between 1% and 1.48%. The main driver of the sawtooth like events is the resonant and non-resonant effect of the (n = 1, m = 3) mode. The instability is stronger for resonant events, and they only appear when β0 = 1.48%. Internal disruptions are mainly driven by the (n = 1, m = 2) mode, and they extend throughout the whole plasma core. Internal disruption events do not show up when resonant sawtooth like events are triggered.

Design of a Vacuum Pumping System for the Closed Helical Divertor for Steady State Operation in LHD

A vacuum pumping system is installed in a Closed Helical Divertor (CHD) in the Large Helical Device (LHD) at the National Institute for Fusion Science for active control of the peripheral plasma density and impurity suppression in the core plasma. In the CHD configuration, the distance between the pumping system and the divertor plates (heat and particle source) is very short (only ∼0.1 m). One of the major issues in designing the pumping system is the reduction of heat load by radiation and thermal conduction due to the neutral particles being released from the heated divertor plates while keeping a high pumping efficiency. Here the heat load and the pumping efficiency are analyzed using a neutral particle transport simulation and a finite element method based software for multi-physics analysis. We propose a new design for a pumping system with an expanded area of the inlet of the water-cooled blinds and a bottom slit beneath the pumping system. This increases the pumping efficiency by approximately 60% over that of our previous design. It also predicts that the increase in heat load on the pumping system for the new design would be reasonably suppressed by a buffer plate with high emissivity on the surface of the vacuum vessel on the inboard side of the torus.

Calculation of the Tangential View of an H-Alpha Intensity Profile in the Large Helical Device

A tangential image of an H-alpha intensity profile in a large helical device (LHD) plasma is investigated by a neutral particle transport simulation using a detailed 3-D grid model including the geometry of the helical plasma and the vacuum vessel. The calculated image of the H-alpha intensity profile calculated by the simulation quite agrees with the observed image, which strongly suggests that there is no significant abnormal neutral particle sources in typical plasma discharge operation in LHD.

Investigation on Total Bremsstrahlung Radiation in High-Density Operation of Large Helical Device Based on Visible Bremsstrahlung Profile Measurement

Visible bremsstrahlung emission profiles have been studied over a wide range of electron densities in H2 gas puff and solid-H2-pellet-fueled discharges of large helical device (LHD). Peaked profiles are observed in high-density discharges (ne≥1014 cm-3) with pellet injection, whereas hollow profiles appeared in the normal discharges (ne≤1014 cm-3) with gas puffing. The total bremsstrahlung radiation is analyzed from the visible bremsstrahlung profile by integrating the energy and plasma volume. It is found that the total bremsstrahlung radiation rapidly increases with the density in the pellet discharges, of which the increment is scaled by the square of density, while it is roughly constant against the density in the gas puff discharges. The total bremsstrahlung radiation becomes equal to the total radiation loss in the pellet discharges. The ratio of the total bremsstrahlung radiation to the total input power has only a range of 3–10% in the gas puff discharges. In contrast, the ratio increases with the density and reaches 30–40% in the pellet discharges. Flat Zeff profiles are observed not only in the gas puff discharges but also in the pellet discharges. This indicates that no impurity accumulation occurs in the high-density operation with H2 pellet injection.

Progress Toward Steady-State Operation in LHD Using Electron Cyclotron Waves

Trials of steady-state operation (SSO) in the Large Helical Device (LHD) were started when a continuous wave (cw) gyrotron with the output power up to 0.2 MW was introduced to the electron cyclotron heating (ECH) system on LHD in 2003. During the first trial of SSO in the seventh LHD experimental campaign in 2004, severe temperature increase on the waveguide transmission line and, as a result, intense pressure increase in the evacuated waveguide occurred, which terminated the operation at 756 s. Additional pumping sections and cooling structures on the transmission line worked well, allowing a 3900-s sustainment of plasma with ne = 0.15 × 1019 m−3 and Te0 = 1.7 keV by 0.1 MW injection power in 2005. The improvement of the ECH system by introducing cw gyrotrons with higher power for further improvement of plasma performance in SSO is in progress.Investigations on electron cyclotron current drive (ECCD) physics have been advanced a few years after the proof of ECCD in LHD. By obliquely injecting second-harmonic X-mode EC waves in toroidal direction, stable current up to 5.5 kA is driven, which was evaluated as a difference in plasma currents of the co- and counter-ECCD discharges with 0.1-MW EC wave power. It takes a few seconds for the driven current to saturate. Change in profile of rotational transform by ECCD and profile of driven current density are investigated by use of motional Stark effect measurement. Peaked and localized driven current profile at the plasma core region was confirmed for on-axis second-harmonic ECCD discharges.

ECRH-Related Technologies for High-Power and Steady-State Operation in LHD

The electron cyclotron resonance heating (ECRH) system on the Large Helical Device (LHD) has been in stable operation for ~11 yr in numerous plasma experiments. During this time, many upgrades to the system have been made, such as reinforcement of the gyrotron tubes, modification of the power supply depending on gyrotron type, and increase in the number of transmission lines and antennas. These efforts allow the stable injection of millimeter-wave power in excess of 2 MW. In parallel, various transmission components were evaluated, and antenna performance was confirmed at a high power level. The coupling efficiency of the millimeter wave from the gyrotron to the transmission line and the transmission efficiency through the waveguide were further improved in recent years. The feedback control of the wave polarization has also been tried to maximize the efficiency of wave absorption. The gyrotron oscillation frequency was reconsidered in order to extend the flexibility of the magnetic configuration in plasma experiments. The development of 77-GHz gyrotrons with the output of 1 MW per few seconds in a single tube is currently taking place in collaboration with the University of Tsukuba. Two such gyrotron tubes already have been installed and were used for plasma experiments recently. An ECRH system with a capability of the steady operation is required, because the LHD can continuously generate confinement magnetic fields using superconducting magnets. Not only the gyrotron but also the transmission system and components must withstand continuous power operation. Further acceleration of both the power reinforcement and a steady-state capability will allow the sustainment of high-performance plasmas.

Interferometer Systems on LHD

This paper describes the interferometer systems on the Large Helical Device (LHD). LHD is equipped with five interferometer systems, each of which has a different operational purpose and measurable electron density range. A single-channel millimeter-wave interferometer is mainly used for low-density plasmas along a horizontal line of sight on the equatorial plane. Wavelengths of 1 and 2 mm are used for vibration compensation based on two-color interferometry, which has been used since the first operation of LHD. A 13-channel CH3OH laser interferometer (wavelength of 119 μm) covers almost the whole poloidal cross sections of LHD plasmas with a chord separation of 90 mm. It routinely provides temporal behavior and profiles of the electron density. The laser has been developed as a collaboration between the National Institute for Fusion Science (NIFS) and Chubu University. An 80-channel CO2 laser interferometer (10.6 μm) is employed for high-density plasmas such as superdense core plasmas. It adopts an imaging technique with three slablike beams and array detectors to measure the density profile precisely. A phase contrast imaging interferometer, which measures density fluctuations, is combined with the CO2 laser interferometer. Since LHD has strong magnetic shear, a distribution of the density fluctuations is evaluated by using shear technique. A conventional millimeter-wave (4 mm) interferometer is also installed at a divertor region to measure dynamic density responses in a divertor leg. The phase counter used on these interferometers was originally developed at NIFS. The phase resolution of a typical phase counter is 1/100 fringe with a temporal response of 10 μs.

DC Power Supplies for LHD Superconducting Magnets and Their Enhancement for Dynamic Control of the Magnetic Field

The Large Helical Device (LHD) has six superconducting coil pairs that form a magnetic field, and plasma experiments of LHD require high accuracy and stability to control the coil current, which must be controlled in a wide range to change the magnetic field. This paper introduces the power system for the LHD superconducting magnets and the enhancement of the power supplies.When designing the power supplies, it is difficult to satisfy the requirements for the current control with a simple controller installed in each power supply because of the tight magnetic coupling between the superconducting coils. In addition, synchronized operation of these power supplies is important for LHD operation. To satisfy the current control requirements and to synchronize operation of the power supplies, six dc power supplies controlled by a computer system were constructed and are operating. The fundamental operation of the LHD is performed under the dc magnetic field, and a power system suitable for steady-state operation was designed and constructed. With the progress in fusion plasma research, more dynamic experiments are planned, and they require dynamic control of the magnetic field. For this purpose, the output voltages of power supplies were enhanced using additional pulse power supplies.

Edge plasma physics/plasma–wall interactions during high density operation in LHD

Superdense plasma with a highly peaked electron density profile was obtained in the reduced recycling discharge in the Large Helical Device (LHD). During the discharge, a core region with a density as high as ∼5 × 10 20 m −3 and a temperature of ∼0.8 keV is maintained by an Internal Diffusion Barrier (IDB) with a steep density gradient. In spite of such a high density at the core region, no serious impurity accumulation has been observed. According to the numerical calculation with the EMC3–EIRENE code, downstream parallel flow in the ergodic region plays an important role to screen the influx of impurities with its strong friction force. The divertor flux distribution and global recycling properties do not change so much as against the change in the dense core plasma.

Significance of MHD Effects in Stellarator Confinement

Substantial progress has been achieved in raising the plasma beta in stellarators and helical systems by high-power neutral beam heating, approaching reactor-relevant values. The achievement of high-beta operation is closely linked with configuration effects on the confinement and with magnetohydrodynamic (MHD) stability.The magnetic configurations of the Wendelstein 7-AS (W7-AS) stellarator and of the Large Helical Device (LHD) and their optimization for high-beta operation within the flexibility of the devices are characterized. A comparative description of the accessible operational regimes in W7-AS and LHD is given. The finite-beta effects on the flux surfaces depend on the degree of configuration optimization. In particular, a large Shafranov shift is accompanied by formation of islands and stochastic field regions as found by numerical equilibrium studies. However, the observed pressure gradients indicate some mitigation of the effects on the plasma confinement, presumably because of the high collisionality of high-beta plasmas and island healing effects (LHD). As far as operational limits by pressure-driven MHD instabilities are concerned, only weak confinement degradation effects are usually observed, even in linearly unstable regimes.The impact of the results concerning high-beta operation in W7-AS and LHD on the future stellarator program will be discussed, including the relationship to tokamak research. Some of the future key issues appear to be the following: the control of the magnetic configuration (including toroidal current control), the modification of confinement and MHD properties toward the low-collisional regime, and the compatibility of high-beta regimes with power and particle exhaust requirements to achieve steady-state operation.

Control, data acquisition and remote participation for steady-state operation in LHD

Control, data acquisition, plasma monitoring and remote participation for steady state operation in the large helical device (LHD) are reviewed. By controlling the impedance matching of ICH, the plasma position and the electron density, high temperature plasma is confined for 1905s. The plasma parameters are monitored in real time. Data are continuously sampled by the YOKOGAWA WE7000 system and by the NATIONAL INSTRUMENTS CompactPCI system. Those data are managed by the object-oriented database system based on ObjectStore in distributed servers with mass storage. By using the multi protocol label switching-virtual private network (MPLS-VPN) technology, the local area network of LHD is expanded to the Japanese fusion community. This provides the remote participants with the same environment of the LHD control room.

Impact of real-time magnetic axis sweeping on steady state divertor operation in LHD

Steady state divertor operation with high performance plasmas (ne ∼ 0.7 × 1019 cm−3, Ti ∼ 2 keV) was demonstrated for half an hour in the Large Helical Device (LHD), the superconducting helical device (R = 3.6–3.9 m, a = 0.6 m, B = 3 T, l/m = 2/10). The high performance plasmas have been sustained with an averaged heating power of 680 kW and achieved an injected energy of 1.3 GJ. This required both advanced technological integration of heating systems and divertor heat flux control. In particular, optimization of divertor heat flux distribution along the divertor leg trace on divertor plates and real-time magnetic axis sweeping (R = 3.67–3.7 m) have allowed LHD to access a steady state regime with a margin of safety for the actively cooled divertor plates. The distribution of divertor heat load along the traces was investigated with calorimetric measurements and it was found that there was a localized heat load connected with the loss of high-energy ions produced by ion cyclotron radio frequency near-fields. Orbit analysis shows that the behaviour of high-energy ions is qualitatively in good agreement with the experimental result. Long-pulse discharges were terminated by radiation collapse due to penetration of metallic flakes into the plasma.

Refueling for Steady-State Plasma by Repetitive Pellet Injection in Large Helical Device

A repetitive pellet injector has been developed for investigation of refueling issues towards the steady-state operation in Large Helical Device (LHD). Continuous operation of more than 10000 pellet launching at 10 Hz has been demonstrated. The maximum repeating rate is 11 Hz. No technical constraint for longer operation has been found. The reliability of pellet launch has exceeded 99.9%. The initial application to the NBI-heated plasmas has been successful in the last experimental campaign of LHD. Although the pulse length is limited by the operational constraint of NBI, the plasma with a density of 8 × 1019 m-3 has been sustained for 2 s by the pellet injection at 10 Hz. A prospect for the future experiment is discussed on the basis of the initial result.

Experiences with Carbon Divertor Operation in LHD

The Large Helical Device (LHD) has employed a graphite helical divertor in a stainless steel vacuum vessel since the beginning of its third experimental campaign in 1999. Initially, installation of the carbon divertor resulted in a dramatic decrease in iron impurities as evidenced by the hollowing out of the radiation profiles and by spectroscopic measurements compared to the previous campaign using a stainless steel divertor. However core radiation and radiation from iron spectral lines gradually increased during the third campaign, which was attributed to sputtering of the stainless steel walls during glow discharge cleaning and redeposition on the carbon divertor tiles.Studies have shown that the graphite tiles were eroded at the divertor leg strike points to a depth of 5 µm during one experimental campaign (~10,000 shots), which is in agreement with numerical calculations using the EDDY code. Carbon is redeposited at and near the strike points along with metallic impurities forming a mixed layer, which may be contributing to iron impurities in the plasma.In LHD radiative collapse resulting from an edge thermal instability caused by radiation of gasses desorbed from the divertor during long pulse discharges terminates the discharge at high density. This radiative collapse is one of the limitations for long pulse discharges up to 150 s.

Engineering prospects of negative-ion-based neutral beam injection system from high power operation for the large helical device

The five-year operation of the negative-ion-based neutral beam injector (N-NBI) system on the large helical device (LHD) is reviewed, and the prospects of negative-ion technology for applying it to future helical fusion reactors are discussed from these experiences. The N-NBI system was designed and constructed based on the results of R&D at NIFS. Using three beamlines, the total port-through injection power of 10.3 MW for 2 s has been achieved. Each beamline attained almost the same performance: ∼3.5 MW of the port-through power with the beam energy of ∼165 keV for hydrogen, which corresponds to the averaged negative ion current density of 25 mA cm−2 from each ion source. Recent modification of accelerators in one of the beamlines was very successful to increase beam energy up to 180 keV and port-through power, 4.4 MW. Long pulse injection heating was also pursued under the steady state LHD operation, and 110 s for 0.1 MW by one ion source, and 80 s for 0.5 MW by two ion sources (one beamline) were achieved. These facts show that the negative ion technology has been established to the same power level of conventional positive ion systems. Several technical improvements needed for obtaining these successful results are described, and the problems that limit the present performance are clarified. Some results of recent R&D to solve these problems are shown.