Search

Experiments on the Large Helical Device with the injection of carbon pellets into discharges of low density have demonstrated a significant reduction of the ion heat conduction in the plasma core and an increase in the central ion temperature by a factor of up to 2. These results are interpreted in the framework of a transport model elaborated on the basis of those applied previously to explain the improvement in confinement by impurity seeding into the tokamak devices TEXTOR and JET. The calculations performed reproduce well the strong peaking of the ion temperature profile with increasing carbon density nZ and the consequent drop in the confinement as nZ exceeds a certain critical level. The importance of different elements in the model, such as braking of the main ion rotation by friction with impurity ones and the shape of the density profiles, are investigated. A qualitative assessment of the applicability under fusion reactor conditions, e.g. of much higher plasma density and heating power, is performed., source:M Z Tokar et al 2020 Plasma Phys. Control. Fusion 62 075008, source:https://doi.org/10.1088/1361-6587/ab8ca2, identifier:0000-0002-9536-9841

In recent deuterium experiments on the large helical device (LHD), we succeeded in expanding the temperature domain to higher regions for both electron and ion temperatures. Suppression of the energetic particle driven resistive interchange mode (EIC) by a moderate electron temperature increase is a key technique to extend the high temperature domain of LHD plasmas. We found a clear isotope effect in the formation of the internal transport barrier in high temperature plasmas. A new technique to measure the hydrogen isotope fraction was developed in the LHD in order to investigate the behavior of the isotope mixing state. The technique revealed that the non-mixing and the mixing states of hydrogen isotopes can be realized in plasmas. In deuterium plasmas, we also succeeded in simultaneously realizing the formation of the edge transport barrier (ETB) and the divertor detachment. It is found that resonant magnetic perturbation plays an important role in the simultaneous formation of the ETB and the detachment. Contributions to fusion reactor development from the engineering point of view, i.e. negative-ion based neutral beam injector research and the mass balance study of tritium, are also discussed., source:https://doi.org/10.1088/1741-4326/ac3cda, identifier:0000-0001-5220-947X

The world's first quasi-axisymmetric stellarator, CFQS, is now under construction. The CFQS will be dedicated to studies on the interaction between flow and turbulence, and confinement improvement by suppression of turbulence in connection with proof-of-principle experiment of quasi-axisymmetry. In order to conduct this experimental research, a heavy ion beam probe (HIBP) system is planned to be installed and utilized to measure the radial electric field and its fluctuation in a CFQS plasma. In this paper, an orbit calculation for a probe beam is performed to verify feasibility of the HIBP in the CFQS. The required beam energy, possible ion species, and the observable region in a CFQS plasma are investigated. The beam attenuation by a CFQS plasma is also estimated for different beam ion species. If we use 133Cs+ as a primary probe beam, the required beam energy is expected to be 30∼50 keV, which is relatively easy to handle. In this case the beam attenuation, evaluated by the ratio between the injected and detected beam currents, is 10−3∼10−2 in a CFQS plasma with a line-averaged electron density of <1.0 × 1019 m−3. For a higher density plasma, usage of 85Rb+ is better in terms of low-beam-attenuation, and a high signal-to-noise ratio. The HIBP in the CFQS will provide a great opportunity to study physics experimentally, related to the radial electric field, poloidal flow, and turbulence suppression., source:A. Shimizu et al 2022 JINST 17 C06004, source:https://doi.org/10.1088/1748-0221/17/06/C06004, identifier:0000-0003-3764-3184

In this paper, the background physics of the isotope effects in the ion internal transport barrier (ITB) are discussed in detail. An heuristic criterion for the ITB strength is defined based on the nonlinear dependence of the ion thermal diffusivity on the local ion temperature in the L-mode phase. Comparing deuterium plasmas and hydrogen plasmas, two isotope effects on the ion ITB are clarified: stronger ITBs formed in the deuterium plasmas and an ITB concomitant edge confinement degradation in the hydrogen plasmas. Principal component analysis reveals that the ion ITB becomes strong when a high input power normalized by the line averaged electron density is applied and electron density profile is peaked. A gyrokinetic simulation suggests that the ITB profile is determined by the ion temperature gradient driven turbulence, while the way the profile saturates in L-mode plasmas is unknown. In the electron density turbulence behavior, a branch transition is observed, where the increasing trend in turbulence amplitude against the ITB strength is flipped to a decreasing trend across the ITB formation. The radial electric field structure is measured by the charge exchange recombination spectroscopy system. It is found that the radial electric field shear plays a minor role in determining the ITB strength., source:T. Kobayashi et al 2021 Nucl. Fusion 61 126013, source:https://doi.org/10.1088/1741-4326/ac298f, identifier:0000-0001-5669-1937

In this study, we developed a calibration method for a fast steering antenna for investigating the mode conversion window used in electron Bernstein wave heating in the large helical device. The calibration was carried out in under-dense plasma against a line-of-sight with an optical thickness which varied spatially. Although multi-reflected background radiation becomes dominant in optically thin lines-of-sight, we succeeded in calibrating the fast steering antenna by including the effect of multi-reflected background radiation in the solution of the radiation transfer equation as the constant by which the temperature of the center of the plasma is multiplied. In addition, we report the initial results of experiments investigating the mode conversion window in over-dense plasma using the calibrated antenna., source:Yuki Goto et al 2019 Jpn. J. Appl. Phys. 58 106001, source:https://doi.org/10.7567/1347-4065/ab3e58, identifier:0000-0001-6636-8500

In this paper, a new criterion for the internal transport barrier (ITB) formation is proposed by defining a unique scalar parameter, the profile gain factor. The profile gain factor shows degree of the confinement improvement with respect to an arbitrary reference temperature profile in the L-mode. As the reference L-mode profile for the large helical device (LHD), the edge ion temperature profile data is extrapolated to the core by the L-mode profile function, which is characterized by the thermal diffusion coefficient being proportional to the local ion temperature. The profile gain factor is defined as the ratio of the ion stored energy experimentally measured to that evaluated from the reference L-mode profile. The proposed method is applied to the LHD experimental data, and its capability for quantification of the ITB strength is demonstrated., source:T Kobayashi et al 2019 Plasma Phys. Control. Fusion 61 085005, source:https://doi.org/10.1088/1361-6587/ab221c, identifier:0000-0001-5669-1937

In this paper, we report new results of a beam switching experiment aiming at a reversed magnetic shear profile formation for the study of an MHD-mode-induced profile collapse event. A transient MHD mode, whose oscillation frequency chirps down, is observed. The electron temperature profile collapse is induced by the mode activity, leading to the flattening of the central electron temperature profile. The radial mode structure is the double-oddparity at the beginning, but it transits to the even-parity in its final stage. The central electron temperature profile recovers after the radial mode structure changes to the even-parity, even though the mode itself does not disappear., source:T. Kobayashi et al 2020 Nucl. Fusion 60 036017, source:https://doi.org/10.1088/1741-4326/ab6b40, identifier:0000-0001-5669-1937, identifier:0000-0003-3754-897X, identifier:0000-0001-7618-6305, identifier:0000-0002-0585-4561

The isotope effect on impurities and bulk ion particle transport is investigated by using the deuterium, hydrogen, and isotope mixture plasma in the Large Helical Device (LHD). A clear isotope effect is observed in the impurity transport but not the bulk ion transport. The isotope effects on impurity transport and ion heat transport are observed as a primary and a secondary effect, respectively, in the plasma with an internal transport barrier (ITB). In the LHD, an ion ITB is always transient because the impurity hole triggered by the increase of ion temperature gradient causes the enhancement of ion heat transport and gradually terminates the ion ITB. The formation of an impurity hole becomes slower in the deuterium (D) plasma than the hydrogen (H) plasma. This primary isotope effect on impurity transport contributes the longer sustainment of the ion ITB state because the low ion thermal diffusivity can be sustained as long as the normalized carbon impurity gradient R/Ln,c, where , is above the critical value (~−5). Therefore, the longer sustainment of the ITB state in the deuterium plasma is considered to be a secondary isotope effect due to the mitigation of the impurity hole. The radial profile of H and D ion density is measured using bulk charge exchange spectroscopy inside the isotope mixture plasma. The decay time of H ion density after the H-pellet injection and the decay time of D ion density after D-pellet injection are almost identical, which demonstrates that there is no significant isotope effect on ion particle transport., source:K. Ida et al 2019 Nucl. Fusion 59 056029, source:https://doi.org/10.1088/1741-4326/ab0e41, identifier:0000-0002-0585-4561

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., source:Y. Takeiri et al 2017 Nucl. Fusion 57 102023, source:https://doi.org/10.1088/1741-4326/aa7fc2, identifier:0000-0001-7569-6717

A feasibility study for installing fast-ion loss diagnostics (FILDs) is initiated to understand energetic ion loss process in a quasi-axisymmetric stellarator CFQS. The calculations of beam ion orbit, the guiding center orbit code including the slowing down by bulk plasma in the Boozer coordinates and the collisionless Lorentz orbit code in the Cartesian coordinates, are performed to estimate the toroidal/poloidal distribution of beam ion losses on a vacuum vessel. We observed that loss points on the vacuum vessel are mainly located at the upper side of torus according to gradient magnetic field drift direction. Based on the loss distribution on the vacuum vessel, a possible location of FILD for CFQS is investigated. Although the energy of lost beam ions is ∼10 keV, relatively high-energy (> 20 keV) components can be measured on the upper side of the torus in the low-density case. Furthermore, the radial or vertical dependence of beam ion loss flux at candidate movable scintillator-type FILD positions is obtained. A considerable number of particles reach the scintillator-type FILD position. The energy and pitch angle distribution of beam ions reaching the FILD position shows that scintillator-type FILDs can measure co-going transit beam ions with energy of ∼25 keV and pitch angle of ∼140 degrees and transition beam ions with energy of 10–15 keV and pitch angle of ∼80–100 degrees., source:Citation K. Ogawa et al 2021 JINST 16 C09029, source:https://doi.org/10.1088/1748-0221/16/09/C09029

The tritium removal from the vacuum vessel is one of the key issues for the realization of a fusion reactor. As in situ tritium removal techniques from the plasma facing materials, helium glow discharge wall conditioning (GDWC) and RF heated plasma due to electron cyclotron wall conditioning (ECWC) and ion cyclotron wall conditioning (ICWC) are applied with Large Helical Device (LHD) which remains tritium produced by deuterium plasma experiment. As the result, it suggests that ICWC could remove the tritium more efficiently than the ECWC and the GDWC. In the ICWC operation, the high energy particles induced by the minority ion heating mode as the standard ICRF heating operation may play an important role in tritium removal from the plasma-facing materials due to the particle's bombardment., source:Citation M Tanaka et al 2021 Phys. Scr. 96 124007, source:https://doi.org/10.1088/1402-4896/ac1bf2, identifier:0000-0001-9941-1958

A remarkable extension of the high-ion-temperature (high-Ti) regime was obtained in deuterium plasma experiments in the Large Helical Device. In order to clarify transport characteristics in the ion internal transport barrier (ITB) formation with an isotope effect, a dataset of pure deuterium (nD/ne > 0.8) and pure hydrogen (nH/ne > 0.8) plasmas in the high-Ti regime were analyzed, and two mechanisms of transport improvement were characterized. A significant reduction of ion heat transport in the core of both deuterium and hydrogen plasmas was observed, indicating ion ITB formation. The dependence of the ion heat diffusivity on temperature ratio (Te/Ti) and normalized Ti-gradient (R/LTi = −(R/Ti)(dTi/dr)) was investigated in the core region, in which gyrokinetic simulations with the GKV code predict the destabilization of ion temperature gradient (ITG) modes. The Te/Ti dependence shows ITG-like property, while a significant deviation from the ITG-like property is found in the R/ dependence, indicating suppression of the ITG mode in the large R/ regime and resultant ion ITB formation. In the comparison between deuterium plasma and hydrogen plasma, the lower transport in the deuterium plasma is observed in both ion and electron heat diffusivities, indicating a significant isotope effect. It was found with the nonlinear turbulent transport simulation with GKV that the zonal flow enhancement contributes to the ITG suppression in the deuterium plasma., source:K. Nagaoka et al 2019 Nucl. Fusion 59 106002, source:https://doi.org/10.1088/1741-4326/ab2d02, identifier:0000-0002-5892-6047

A simultaneous high ion temperature (Ti) and high electron temperature (Te) regime was successfully extended due to an optimized heating scenario in the LHD. Such high-temperature plasmas were realized by the simultaneous formation of an electron internal transport barrier (ITB) and an ion ITB by the combination of high power NBI and ECRH. Although the ion thermal confinement was degraded in the plasma core with an increase of Te/Ti by the on-axis ECRH, it was found that the ion thermal confinement was improved at the plasma edge. The normalized ion thermal diffusivity ${{\chi}_{\text{i}}}/T_{\text{i}}^{1.5}$ at the plasma edge was reduced by 70%. The improvement of the ion thermal confinement at the edge led to an increase in Ti in the entire plasma region, even though the core transport was degraded., source:H. Takahashi et al 2017 Nucl. Fusion 57 086029, source:https://doi.org/10.1088/1741-4326/aa754b, identifier:0000-0001-6984-9174

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., source:H. Takahashi et al 2018 Nucl. Fusion 58 106028, source:https://doi.org/10.1088/1741-4326/aad87e

The long pulse experiments in the Large Helical Device has made progress in sustainment of improved confinement states. It was found that steady-state sustainment of the plasmas with improved confinement at the core region, that is, electron internal transport barrier (e-ITB), was achieved with no significant difficulty. Sustainment of a plasma having e-ITB with the line average electron density ne_ave of 1.1 × 1019 m−3 and the central electron temperature Te0 of ∼3.5 keV for longer than 5 min only with 340 kW ECH power was successfully demonstrated., source:Citation Y Yoshimura et al 2018 Plasma Phys. Control. Fusion 60 025012, source:https://doi.org/10.1088/1361-6587/aa9950, identifier:0000-0001-6744-1829

Using ion cyclotron heating and electron cyclotron heating (ECH), or solo ECH, trials of steady state plasma sustainment have been conducted in the superconducting helical/stellarator, large helical device (LHD) (Ida K et al 2015 Nucl. Fusion 55 104018). In recent years, the ECH system has been upgraded by applying newly developed 77 and 154 GHz gyrotrons. A new gas fueling system applied to the steady state operations in the LHD realized precise feedback control of the line average electron density even when the wall condition varied during long pulse discharges. Owing to these improvements in the ECH and the gas fueling systems, a stable 39 min discharge with a line average electron density ne_ave of 1.1 × 1019 m−3, a central electron temperature Te0 of over 2.5 keV, and a central ion temperature Ti0 of 1.0 keV was successfully performed with ~350 kW EC-waves. The parameters are much improved from the previous 65 min discharge with ne_ave of 0.15 × 1019 m−3 and Te0 of 1.7 keV, and the 30 min discharge with ne_ave of 0.7 × 1019 m−3 and Te0 of 1.7 keV., source:Citation Y. Yoshimura et al 2016 Nucl. Fusion 56 046005, source:https://doi.org/10.1088/0029-5515/56/4/046005, identifier:0000-0001-6744-1829

The central electron temperature has successfully reached up to 7.5 keV in large helical device(LHD) plasmas with a central high-ion temperature of 5 keV and a central electron density of1.3×1019 m−3. This result was obtained by heating with a newly-installed 154 GHz gyrotronand also the optimisation of injection geometry in electron cyclotron heating (ECH). Theoptimisation was carried out by using the ray-tracing code ‘LHDGauss’, which was upgradedto include the rapid post-processing three-dimensional (3D) equilibrium mapping obtainedfrom experiments. For ray-tracing calculations, LHDGauss can automatically read the relevantdata registered in the LHD database after a discharge, such as ECH injection settings (e.g.Gaussian beam parameters, target positions, polarisation and ECH power) and Thomsonscattering diagnostic data along with the 3D equilibrium mapping data. The equilibrium mapof the electron density and temperature profiles are then extrapolated into the region outsidethe last closed flux surface. Mode purity, or the ratio between the ordinary mode and theextraordinary mode, is obtained by calculating the 1D full-wave equation along the directionof the rays from the antenna to the absorption target point. Using the virtual magnetic fluxsurfaces, the effects of the modelled density profiles and the magnetic shear at the peripheralregion with a given polarisation are taken into account. Power deposition profiles calculatedfor each Thomson scattering measurement timing are registered in the LHD database. Theadjustment of the injection settings for the desired deposition profile from the feedbackprovided on a shot-by-shot basis resulted in an effective experimental procedure., source:Citation T. Ii Tsujimura et al 2015 Nucl. Fusion 55 123019, source:https://doi.org/10.1088/0029-5515/55/12/123019, identifier:0000-0002-2983-5920

A real-time interlock system for power injection in electron cyclotron resonance heating(ECRH) was developed to be applied to Large Helical Device (LHD) plasma. This systemenabled perpendicular injection, thus improving the performance of ECRH more than has everbeen achieved before in LHD. Perpendicular propagation of the electron cyclotron wave at77 GHz became more insensitive to the effect of refraction in comparison to the conventionaloblique propagation. The achieved central electron temperature in the case of perpendicularinjection was approximately 2 keV higher than that in the case of standard oblique injectionfor a central electron density of 1 × 1019 m−3 by 1 MW injection.With such improvedperformance of ECRH, high-density ECRH plasma of 8 × 1019 m−3 was successfullysustained after the injection of multiple hydrogen ice pellets for the first time in LHD., source:Citation T.I. Tsujimura et al 2021 Nucl. Fusion 61 026012, source:https://doi.org/10.1088/1741-4326/abc977, identifier:0000-0002-2983-5920

Collective Thomson scattering (CTS) is one of attractive diagnostics for measuring locally and directly the fuel temperature and the velocity distribution of fast ions in fusion plasmas. A mega-watt class source of millimeter or sub-millimeter waves is required to detect a weak scattered radiation superimposed on background radiation owing to electron cyclotron emissions (ECEs) from plasmas. Based on electron cyclotron resonance heating (ECRH) system with the frequencies of 77 GHz and 154 GHz in the Large Helical Device (LHD), the CTS diagnostic system has been developed to measure bulk ion temperatures from a few keV to ∼10 keV and fast ions originated from 180 keV-neutral beam injection in the LHD. The measured CTS spectra and their time evolutions are analyzed with the electrostatic scattering theory. The bulk ion temperatures obtained from CTS spectra increase with the neutral beam injections and decrease with the heating terminated. The velocity map of simulated fast ions explains that the bumps on tail of measured CTS spectra are caused by the co- and counter- fast ions. A new prescription for anisotropic velocity distribution function is proposed. As for 154 GHz bands, the CTS spectrum broadenings for D and H plasmas are distinguished reasonably at the same temperature, and its ion temperatures are comparable to those of the charge exchange recombination spectroscopy. As reactor-relevant diagnostics, a 300 GHz gyrotron and a corresponding receiver system have been implemented in LHD to access high density plasmas with low background ECEs. The recent progress for CTS diagnostics and their spectrum analysis with the probe frequencies of 77 GHz, 154 GHz, and 300 GHz in the LHD experiments is described., source:M. Nishiura et al 2020 JINST 15 C01002, source:https://doi.org/10.1088/1748-0221/15/01/C01002, identifier:0000-0003-2410-7550

Detection of cell-free DNA (cfDNA) has an impact on DNA analysis in liquid biopsies. However, current strategies to detect cfDNA have limitations that should be overcome, such as having low sensitivity and requiring much time and a specialized instrument. Thus, non-invasive and rapid detection tools are needed for disease prevention and early-stage treatment. Here we developed a device having a microheater integrated with zinc oxide nanowires (microheater-ZnO-NWs) to detect target single-stranded DNAs (ssDNAs) based on DNA probe hybridization. We confirmed experimentally that our device realized in-situ annealed DNA probes by which we subsequently detected target ssDNAs. We envision that this device can be utilized for fundamental studies related to nanobiodevice-based DNA detection.

0000-0001-5669-1937, In this paper, the background physics of the isotope effects in the ion internal transport barrier (ITB) are discussed in detail. An heuristic criterion for the ITB strength is defined based on the nonlinear dependence of the ion thermal diffusivity on the local ion temperature in the L-mode phase. Comparing deuterium plasmas and hydrogen plasmas, two isotope effects on the ion ITB are clarified: stronger ITBs formed in the deuterium plasmas and an ITB concomitant edge confinement degradation in the hydrogen plasmas. Principal component analysis reveals that the ion ITB becomes strong when a high input power normalized by the line averaged electron density is applied and electron density profile is peaked. A gyrokinetic simulation suggests that the ITB profile is determined by the ion temperature gradient driven turbulence, while the way the profile saturates in L-mode plasmas is unknown. In the electron density turbulence behavior, a branch transition is observed, where the increasing trend in turbulence amplitude against the ITB strength is flipped to a decreasing trend across the ITB formation. The radial electric field structure is measured by the charge exchange recombination spectroscopy system. It is found that the radial electric field shear plays a minor role in determining the ITB strength.