In this paper, a method for estimating the radial profile of electron density n_e using a single line-of-sight signal by the He I line intensity ratio method is proposed. By applying this method to cylindrical helium plasma,in which electron temperature was almost uniform and density was uniform in the center, we tried to estimate the parameters representing spatial distribution. It was confirmed that a good distribution estimation result could be obtained by considering the sensitivity factor, the rate at which the line intensity ratio changes as the parameters change, during optimization. Two methods of considering a sensitivity factor are proposed: using the best combination of intensity ratios for analysis in terms of the sensitivity factor, and weighting the objective function using the sensitivity factor. The former method can be analyzed in short computational time, although its applicability is limited. The latter method can be used when it is not obvious which set of intensity ratios is best to use, although it takes more computational time compared with the former method. Both methods reproduce the parameter of a radial density profile.

This study investigated the chemical distribution of an atmospheric-pressure plasma jet (APPJ) along its propagation direction using numerical simulation. Low-resolution spectral data were used to estimate the gas temperature and the excitation temperature. These estimations were used with a collisional-radiative model to elucidate population densities and the electron temperature. A global model was applied to investigate the chemical species distribution in the plasma jet. The thermodynamic properties of the APPJ corresponded well to the relation T_g < T_exc < T_e for all the positions along the jet propagation. Chemical species generation and propagation along the plasma jet were numerically simulated using the GM with input parameters derived from the CR model and the ideal gas law.

We report herein an automated scheme for generating a mesh based on a structure of magnetic-field lines. For all angles in the toroidal subdomain, the mesh generated by this scheme must not contain broken tetragons. To satisfy this requirement, we propose herein a first method to detect broken elements and a second method to automatically modify the broken elements. We use the proposed mesh-generation scheme with magneticfield data acquired on the magnetic axis of the Large Helical Device at R_ax = 3.75 m. The results show that broken elements are detected by the first method, and the second method transforms broken elements into regular elements.

Extreme ultraviolet (EUV) and vacuum ultraviolet (VUV) wavelength spectra including line emissions released from neon (Ne) ions ranging from low to high charge states observed simultaneously in a single discharge are summarized for contribution to compile a fundamental spectral dataset for the Ne-seeded divertor heat load reduction experiments in Large Helical Device (LHD). NeIX and NeX lines were observed in the EUV wavelength range of 10∼50 Å and NeIII-NeVIII lines were observed in the VUV wavelength range of 400∼1000 Å. The temporal evolutions of the line intensities exhibited different behaviors between the edge emissions of NeIII-NeVIII with the ionization potential, E_i, of 63∼239 eV and the core emission of NeX with E_i of 1362 eV. NeIX with E_i of 1196 eV exhibited a marginal behavior between the edge emission and the core emission.

The total neutron emission rate (S_n) characteristics of electron cyclotron heated plasma were surveyed in the Large Helical Device in order to exhibit the thermonuclear performance of helical plasma. The dependence of S_n on electron density showed that S_n increased with an electron density of power of 3.1. To understand S_n, characteristics in the electron cyclotron heated plasma, a numerical simulation considering thermal deuterium-deuterium fusion reactions was performed. Although the numerical simulation overestimated S_n in a relatively low S_n region, calculated S_n matched the experimental result for a relatively high S_n region. A possible reason for the disagreement in the low S_n region is that effective charge due to the impurities such as carbon is changed because of the low density.

A radiative collapse predictor has been developed using a machine-learning model based on high-density plasma experiments in the Large Helical Device (LHD). Concurrently, the physical background of radiative collapse was discussed based on the distinct features extracted by a sparse modeling, which is one of the frameworks of data-driven science. Electron density, CIV and OV line emissions, and electron temperature at the plasma edge have been extracted as the key parameters of radiative collapse. Those parameters are relevant to the physical knowledge that the major cause of radiative collapse is the enhancement of radiative loss by light impurities in the plasma-edge region. Using these four parameters, the likelihood of occurrence of radiative collapse has been estimated. The behavior of plasma at the edge—in particular, the carbon impurities outside the last closed flux surface—has been evaluated using EMC3-EIRENE code for the phase with increasing likelihood, that is, the plasma is getting close to the collapse. It is shown that the radiation caused by the C³⁺ ion, which corresponds to the CIV emission, is enhanced in the region where electron temperature is around 10 eV.

A novel divertor biasing using four biasing plates that are arranged toroidally every 90^◦ on the upper divertor plate is applied to low-density plasmas of the QUEST spherical tokamak. When some of these plates are biased in-phase by applying a sawtooth waveform voltage of 85-V amplitude and 50-Hz repetition, up to approximately 35% reduction of the particle flux to the divertor is observed during positive biasing. The input power for the flux reduction is approximately 0.2 kW for low-density tokamak plasmas produced by ∼130-kW electron cyclotron wave injection. Additionally, the signal of a plate probe placed in the low-field side of the mid-plane of the vacuum vessel indicates enhanced losses of fast electrons during positive biasing. The enhanced loss is attributed to small resonant magnetic perturbations produced by the bias-driven currents in the scrape-off layer. This novel divertor biasing is expected to provide a new experimental tool for studying divertor heat load control and fast electron confinement in a tokamak device.

Polarization of the hydrogen Lyman-α line is detected in the Large Helical Device. It is the first observation of a polarized atomic emission line in magnetically confined fusion plasma devices. With the help of an atomic model simulation, the anisotropy in the electron velocity distribution function (EVDF) in terms of T_∥/T_⊥ is evaluated, where T_∥ and T_⊥ represent the electron temperature in the parallel and perpendicular directions regarding the magnetic field, respectively. The results show that T_∥/T_⊥ has a tendency to decrease and deviate from unity with decreasing electron-electron collision frequency, which qualitatively agrees with an intuitive understanding of the anisotropic EVDF in the plasma boundary.

A plasma density fluctuation signal was measured by the recently-installed Beam Emission Spectroscopy (BES). A time-dependent analysis was performed for a discharge in the Large Helical Device (LHD) and compared with the magnetic fluctuation. While the fundamental frequency peak shows a high correlation between the density fluctuation and the magnetic fluctuation, the higher harmonic components have smaller or even negligible correlation. As a possible mechanism that makes the density fluctuation and the magnetic fluctuation different, the relation between the MHD mode and the transport is discussed.

In a small tokamak, the visible light emission is observed and used to investigate plasmas' behavior with a fast visible camera. However, the reflected light causes a systemic error in measuring visible light emitted from the plasma. In this paper, we managed to overcome the reflection effect with the ray-tracing technique which is utilized in a synthetic diagnostic platform of the small tokamak device PHiX at Tokyo Institute of Technology using Raysect and CHERAB python libraries. We successfully evaluated the amount of reflected light and obtained tomographic reconstruction images from simulated and experimental data with the Tikhonov-Phillips regularization and the L-curve method to choose an optimal regularization parameter. We also proposed to project the contour of a reconstruction image onto a camera image to validate tomography results.

The controllability of particle supply during long-term discharge in a high-temperature environment was investigated at the Q-shu University Experiment with steady state spherical tokamak (QUEST). QUEST has a high-temperature wall capable of active heating and cooling as a plasma-facing wall. With this hot wall, a temperature rise test was conducted with 673 K as the target temperature. It was confirmed that the hot wall could maintain the temperature above 600 K. Feedback control of particle fueling was conducted to control Hα emission, which is closely related to influx to the wall. Using this particle fueling control and setting the hot wall temperature to 473 K, it was possible to obtain a discharge of more than 6 h. In this discharge, the fueling rate of particles decreased with time, and finally became zero, losing the particle fueling controllability. However, as soon as the cooling water started to flow through the hot wall, particles could be supplied again, and controllability was restored. Thus, indicating that temperature control of the plasma first wall is important even in the high-temperature environment of 473 K to control particle retention of the wall.

Influences of nitrogen ratio on plasma detachment and molecular activated recombination (MAR) processes during combined seeding with hydrogen have been investigated utilizing end-loss plasma in the GAMMA 10/PDX tandem mirror. Additional gases were injected under the condition that hydrogen partial pressure was fixed and nitrogen partial pressure was changed from 0% - 10% compared to that of hydrogen. Electron density and ion flux further decrease with increasing nitrogen ratio. In addition, it is suggested that the hydrogen-MAR process that begins with dissociative attachment is suppressed during combined seeding of nitrogen and hydrogen. Observed emission spectrum of NH radicals suggests that the density of NH increases as nitrogen ratio increases and nitrogen-induced MAR efficiently contributes to the reduction of particle flux.

The behavior of energetic particles (EPs) associated with toroidal Alfvén eigenmode (TAE) activities during the combined injection of hydrogen and deuterium beams was investigated in the Large Helical Device (LHD). The enhanced transports of both proton and deuteron with TAE activities were simultaneously observed by a tangentially viewing and mass and energy resolved neutral particle analyzer (E||B-NPA). At the timing of the TAE bursts with the mode number n = 1, both proton and deuteron were transported to the outboard and observed with the similar energies of 137 - 138 keV. At the peak amplitude of the magnetic fluctuations measured by the Mirnov coils, the mixed frequencies of 64 kHz and 29 kHz were identified, and the observed frequencies did not chirp down. The observed timings of the transported hydrogen and deuterium were just after the magnetic fluctuations of 64 kHz and 29 kHz, respectively. By adapting the cross-correlation analysis, the delay times from the magnetic fluctuation to the detection of EPs by E||B-NPA are estimated to 95 µs and 145 µs for hydrogen and deuterium, respectively. These delays are considered to be the time of the radial transport, and the time delays depended on the velocities of the transported EPs.

GAMMA 10/PDX is a linear plasma confinement device that mainly uses slow waves in ion cyclotron range of frequencies (ICRF) for heating, and it can achieve a high ion temperature that is in the range of several keV. Although slow waves are effective for ion heating, it is difficult to excite them with antennas at high densities, such as those over 10¹⁹ m⁻³. In this study, we considered a way to excite a slow waves using the frequency difference of input waves. Two fast waves having different frequencies, which are both adequate for high-density plasma, were applied with antennas to excite a slow wave as a difference-frequency wave between the fast waves. As a result, the excitation of difference-frequency waves inside the plasma was confirmed from magnetic probe and reflectometry measurements, and an increase in diamagnetism was also observed. These results firstly demonstrate the possibility of slow-wave heating using a DF wave in a high-density linear plasma, where direct slow-wave heating is not feasible.

Transient coaxial helicity injection (t-CHI) current start-up using a new design simple electrode configuration has been implemented on the QUEST. Discharges injected from the low field side (LFS) and from the high field side (HFS) were examined. Compared to the LFS injection case, the HFS injection has the advantages of providing access to a higher toroidal field and better controlling the location of the injector flux footprint location. Although the present PF coils on QUEST are not well positioned to form the injector flux on the HFS injector region and there has been a frequent occurrence of the spurious arcs, known as absorber arcs, HFS injection has shown flux evolution in a shape that is suitable for the formation of closed flux surfaces. The discharges were improved by installing an in-vessel-coil and adding a new cylindrical electrode to the existing CHI electrode. The results show that the new cylindrical electrode allowed the flux to evolve stably while allowing both the inner and the outer injector flux footprint to remain in the vicinity of the cylindrical electrode. This configuration which inherently generates a narrow injector flux footprint width resulted in discharges that strongly suggested the persistence of the CHI generated plasma after the injector current was reduced to zero. These studies have informed us of the need to improve the CHI gas injection system so that the absorber arcs could be better controlled in the HFS injection configuration.

Energy flow in the collisional merging process of a field-reversed configuration (FRC) was experimentally evaluated. Collisional merging formation of an FRC was carried out in the FAT-CM (FRC amplification via translation-collisional merging) device. In this experiment, the field-reversed theta-pinch formed FRC-like plasmoids are accelerated due to a magnetic pressure gradient. Then, two plasmoids collide at a relative speed of ∼300 km/s, which is faster than typical Alfvén and ion sound speeds (∼50 km/s) on the separatrix. The kinetic and internal energy of plasmoids before and after collision are estimated by simultaneous multi-point measurements combining magnetic probes and interferometers. The energy flow in the collisional merging process is compared to an experimental case with single plasmoid translation. This comparison indicates that the kinetic energy of two accelerated plasmoids regenerates back into the internal thermal energy of the FRC after merging. Moreover, density and neutron measurements suggest excitation of shockwaves. These results indicate that shock heating may become a channel for energy regeneration.

Neutral beam (NB) fast-ion loss mechanisms in the large helical device (LHD) are investigated by the combination of the neutron measurement, the classical slowing-down simulation, and the neo-classical guiding center orbit following simulation. It is found that the neo-classical transport provides little contribution to the loss of tangentially injected NB fast-ions. For perpendicularly injected NB fast-ions, the neo-classical transport has more than 40% contribution to the NB fast ion loss. These results indicate that there are other loss mechanisms dominant in LHD plasmas. The charge exchange loss is one of the plausible candidates for the loss mechanism.

Fuel particle and impurity influxes have been investigated for ADITYA-U tokamak plasma operated with toroidal belt limiter using PMT based spectroscopic diagnostic system installed on machine. The influxes of hydrogen and impurity ions are estimated using various lines of sight (LoS) terminating on the graphite limiter and stainless steel wall to understand their contributions in recycled particle and impurities into the main plasma. It is found that the influxes of neutral hydrogen and oxygen are around 4 times higher in case of LoS terminating on the limiter than the wall while carbon influxes from the both LoSs are comparable. The comparable integrated particle influxes from both LoSs indicate the important role of the wall in the recycling and presence of the impurities in the plasma. The particle confinement time (τ_p) and recycling coefficient (R) are also estimated to quantify those from the estimated particle influxes. The τ_p values vary between 8 to 25 ms when plasma electron density is in the range of 2.0 - 3.2 × 10¹⁹ m⁻³. Analysis of recycling coefficient, R suggests that the Plasma Facing Component (PFC) acts as the particle sink at the beginning of the plasma operational campaign. The R values tend to become more than one as the campaign progresses suggesting that the PFC acting as the particle source.

The Steady state operation can be defined for various characteristic physical time scales. Usually one speaks of steady state plasmas when the discharges are stationary for several energy confinement times. The next longer time constant is the L/R time for the development of the toroidal plasma current. Ultimately, the gas equilibrium must also be achieved. W7-X is ideal for creating stationary conditions for the different plasma parameters. In the operation phase OP1.2, the experimental operation was neither limited by the confining magnetic field nor by the pulse length of the ECRH. Only the maximum tolerable temperature of the uncooled in-vessel components and the test divertor unit (TDU) limited the total input energy. Therefore this paper presents steady state operation different plasma parameters with different time scales.

A stationary direct current of a central solenoid (DC_CS) can expand the operation region of tokamak plasmas. For tokamak plasma formation experiments using electron cyclotron heating (ECH) alone, the minimum ECH power and plasma current necessary for tokamak plasma formation were reduced using the DC_CS, which was applied in the counter rotational direction to the plasma current. In TST-2, the minimum ECH power necessary for the formation of a tokamak plasma was reduced from 3.3 to 1.6 kW when the DC_CS was changed from 0 to 184 A. Simultaneously, the plasma current that was needed to sustain the tokamak plasma configuration was reduced to 0.6 kA.

The physical processes of ion heating during high guide field (B_g/B_rec ∼ 6) reconnection is explored utilizing Doppler tomography measurement in the TS-6 ST (Spherical Tokamak) merging experiment. The newly developed 288CH extensive/high-resolution (Δr ∼ 1.5 cm; ΔZ = 1.0 cm) ion Doppler spectroscopy system has revealed the detailed characteristic of ion heating in the downstream region. In the high guide field regime,ion heating occurs not only around the diffusion region but also more globally in the downstream region. This was beyond the scope of the previous ion temperature measurement and hence we have confirmed a different heating mechanism which is attributed to strong in-plane electric field produced extensively in high guide field reconnection regime.

Full vertical profiles of visible bremsstrahlung continuum have been measured for diagnostics of effective ion charge, Z_eff , with parallel optical fiber arrays at horizontally elongated plasma cross section in the Large Helical Device (LHD). Measured bremsstrahlung profiles were asymmetric due to non-uniform bremsstrahlung emissions located in the stochastic magnetic field layer. Then, the Z_eff profile analysis has been done only for a magnetic configuration with relatively thin stochastic magnetic field layer, i.e. for plasmas at magnetic axis position of R_ax = 3.60 m. The effect of the non-uniform bremsstrahlung emission disappears in the lower half profile at R_ax = 3.60 m, while the Z_eff profile analysis is extremely difficult in other magnetic configurations due to the presence of the non-uniform bremsstrahlung emission over the entire vertical profile from top to bottom. To improve the difficult situation in the Z_eff diagnostic the fan array optical fiber system was newly installed at vertically elongated plasma cross section for full horizontal profile measurement. It is found that the non-uniform bremsstrahlung emissions disappear from the signal and the observed bremsstrahlung profile is almost symmetric between inboard and outboard radial profiles. The analysis on the Z_eff profile has been carried out based on the plasma equilibrium database in LHD, TSMAP. A preliminary result is obtained in high-density discharges, which shows a flat Z_eff profile and values of Z_eff ∼ 1.

LHDGauss is a multi-ray tracing code to calculate microwave beam propagation and power deposition of electron cyclotron heating (ECH) using the electron density and electron temperature profiles in the Large Helical Device (LHD) plasma. LHDGauss also takes into account the injected wave polarization purity. LHDGauss uses the cold plasma dispersion to calculate ray propagation and derives the power absorption profile of ECH by using the absorption coefficient based on weakly relativistic plasmas and the wave propagating perpendicularly to the magnetic field. However, the ECH beam propagation is mainly oblique to the magnetic field in the LHD and it is necessary to take into account the angular dependence of the absorption coefficient in order to derive more accurate ECH power deposition profiles. We upgraded LHDGauss and were able to calculate the absorption coefficient using the weakly relativistic dielectric tensor which considers the influence of the angle between the magnetic field and the wave vector. We compared the power deposition profiles calculated by previous and upgraded LHDGauss with the changes of the electron temperature profile of the LHD plasma in the oblique and the perpendicular O1-mode ECH injection cases. We have obtained more reasonable power deposition profiles by using the upgraded LHDGauss in both the perpendicular and the oblique injection cases.

A tomographic method using tangential visible light is proposed for MHD equilibrium reconstruction via two processes. The first process is a tomographic method to estimate the last-closed-flux-surface (LCFS) in two-dimensional poloidal cross-section using a single tangential camera image. Applying the Laplacian eigenfunction series expansion and L₁ regularization, we can reconstruct the LCFS from relatively sparse and noisy observations. The second method is a free-boundary tokamak equilibrium calculation using the TASK/EQU code, in which we use the estimated plasma surface information as the constraints for the equilibrium calculation. As a result, we develop a new method for identifying equilibrium states using visible light information.

For accelerating a transport simulation with an advanced physics turbulent transport model like TGLF, we have been developing a surrogate model that mimics the behavior of the model based on a neural network model. With a steady-state transport solver GOTRESS used, the surrogate model has shown its ability to successfully predict temperature profiles almost equivalent to those by TGLF. The performance of the surrogate model is improved by optimizing hyperparameters and eliminating outliers from training data. Extrapolability of the optimized model is examined by changing the normalized temperature gradient. The objective is to better investigate the nature of the model in addition to measuring its utility in transport simulations. The versatile model, which has been trained with data of multiple cases, is developed applicable to many situations. It shows the same reproducibility as the model specific to each individual case, a fact which unveils great potential of the surrogate model in transport simulations.

Long pulse discharges in the Large Helical Device have often been interrupted by large amounts of dust particle emission from the divertor region caused by the exfoliation of carbon-rich mixed material deposition layers. The plasma wall interaction code ERO2.0 has provided the simulation results of the three-dimensional distribution of the carbon flux density in the divertor region which is quite reasonable with the observed distribution of the carbon-rich deposition layers. The code has also succeeded in reproducing the reduction of the carbon deposition layers on dome plates by changing the target plate configuration in the divertor region. The ERO2.0 simulations have also successfully explained dust particle emission from the inboard side near the equatorial plane for the new target plate configuration at the termination of a long pulse discharge. These simulation results prove that the ERO2.0 code is applicable to predicting the possible position from where the dust particles are released, and to designing an optimized divertor configuration for performing stable long pulse discharges with controlled dust particle emission.

The ensemble Kalman smoother (EnKS) is introduced to the data assimilation system, ASTI, based on the integrated transport simulation code, TASK3D. We use the EnKS to estimate state variables composed of electron and ion temperature, density, and numerical factors of turbulent transport models and neutral beam injection (NBI) heat deposition. The time series data of plasma temperature and density profiles are assimilated into TASK3D. The estimation performance of the EnKS is investigated, and the EnKS is applied to an NBI plasma in the Large Helical Device (LHD) (shot:114053) to estimate the factors of the turbulent heat transport model. The obtained factors can reproduce the experimental temperature data with high accuracy. These results indicate the effectiveness and validity of the EnKS approach for accurate estimation of fusion plasma parameters.

The better radial transport modelling of the SOL/divertor plasmas is required to obtain quantitative agreements between experimental values and simulation results towards the proper SOL/divertor plasma prediction of the future magnetic fusion devices. Various SOL/divertor codes (e.g. SONIC, SOLPS-ITER, EDGE2D, UEDGE, EMC3-Eirene, etc.) have been developed and applied to the interpretative simulations of the present fusion devices. Above codes treats the radial transport as a diffusion model and there exists uncertainty to decide the value of the diffusion coefficients. The simulation by the SONIC has resulted in a few times larger ion flux towards the divertor plate than the experimental value observed in JT-60U possibly due to such an uncertainty of the diffusion process. To achieve the quantitative agreement between the experiments and the simulation results is indispensable to predict SOL/divertor plasma parameters of the future fusion devices. Here, radial transport was simulated by considering large-angle elastic scattering between ions and neutral particles and by applying the probability distribution function to fluid equations in the divertor plasma. It was found that the density peak and ion flux are reduced by about half.

Toroidal torque by electron cyclotron heating (ECH) is investigated in the Large Helical Device (LHD) plasmas, assuming the supra-thermal electrons by ECH generate torques on the plasma through j_r × B and collisions. The j_r × B torque depends on the radial drift velocity and the fraction of trapped electrons. Therefore, the magnetic configuration and the heating location affects the toroidal torque. We investigate the magnetic configuration and heating location dependences of toroidal torques by ECH in LHD, by considering three typical magnetic configurations: the inward shifted, standard, and outward shifted configurations. As a result, magnetic ripple bottom heating generates larger torque than that of ripple top heating because of the large fraction of trapped electrons. Also, heating at the outer minor radius generates larger toroidal torque than that of heating at the inner radius, and the injection angle can also change the toroidal torque profile. Moreover, ECH generates the largest toroidal torque in the outward shifted configuration. Finally, we evaluate the toroidal flow velocities with the obtained toroidal torques. We obtained the largest flow near the axis in the standard configuration because of its small viscosity and large toroidal torque.

The characteristics of the magnetohydrodynamics (MHD) instabilities for low edge safety factor (low-q) plasma are investigated on J-TEXT with three-dimensional (3D) nonlinear MHD Infrastructure for Plasma Simulation (MIPs) code. The dynamics of m = 2 mode coupling with m = 1 mode is found in the linear growth phase with the initial q_axis < 1 equilibrium. And new 3D helical coils are designed for generating additional rotational transform based on the J-TEXT configuration, which is helpful to suppress the amplitude of the MHD instabilities from the modeling result.

In this research, we investigate the effect of magnetic flux tube expansion on the divertor plasma parameters by using the fluid code “LINDA”. A comparison between the cylindrical flux tube (without the magnetic flux expansion) and the expansion magnetic flux tube has been undertaken. The aim of the study is to understand the impact of magnetic field expansion on the divertor physics by using the LINDA fluid code. The plasma density (n_i) is decreased and parallel velocity is increased (u_i||) toward the target plate with the expansion of magnetic field lines near the target plate. The heat and particle fluxes are reduced significantly on the target plate in the case of the expansion mesh configuration. For the case of cylindrical mesh, advection becomes stronger with the decreasing distance from the target plate. In the case of expansion mesh, diffusion is stronger with the decreasing distance from the target plate. These outcomes clearly indicate the effect of the magnetic field structure on the divertor plasma parameters.

We develop novel numerical schemes for electrostatic field solvers toward the whole-volume gyrokinetic simulation of stellarators. The gyrokinetic Poisson equation should be solved for complicated magnetic fields in the stellarator without assuming nested flux surfaces and toroidal symmetry. The developed schemes enable us to generate suitable unstructured meshes and obtain the solutions within a limited numerical cost for general magnetic field structures. These schemes will be integrated and utilized in X-point Gyrokinetic Code - Stellarator (XGC-S).

Distribution of electric potential near the wall in the volume produced negative ion source with the magnetic field increasing toward the wall such as the cusp magnetic field is investigated analytically. The plasma-sheath equation that gives the electric potential in the plasma region and the sheath region near the wall is derived analytically and the potential distribution near the wall is obtained by solving the plasma-sheath equation. Effects of the degree of increase of the magnetic field, the production amount of volume produced negative ion, and the ion temperature on the distributions of electric potential are shown.

We investigated radiation distributions, fluences, generation rates, and radioactive isotopes induced by heavy-ion impact into a vacuum vessel wall for a particle accelerator in the heavy-ion inertial fusion (HIF). Numerical results showed that the radiation management should be considered for neutrons and photons during the HIF system operation. Compared to the detailed numerical simulation, estimations in total reaction cross-sections and threshold kinetic energy for interaction in the Coulomb barrier were expected to provide criteria as a safer side for radiation management in the HIF accelerator. The activation and the isotopes generated by the interaction between the heavy-ion and the wall material were obtained from the calculation results, which showed that the generation rate and the variation of the generated isotope increased with the incident kinetic energy of the heavy-ion beam. The calculation results also indicated that the management of radioactive materials is important for the HIF accelerator system's safe operation.

An electron spectrometer was used to measure the electron and ion spectra in two different irradiations of implosion laser to the deuterated spherical shell targets at the direct fast ignition experiments on the Gekko-LFEX facility. In the transverse irradiation against the LFEX laser axis, the low effective hot electron temperature (T_eff) and huge neutron yield (Ny) could be obtained although high T_eff and low Ny could be observed in the axial irradiation. In the transverse irradiation, the efficient core heating could be obtained because the laser-plasma interaction position is close to the core and the diffusive/ion drag heating may be effective.

Two types of acceleration techniques, H-matrix arithmetics and an H-matrix-based variable preconditioning (VP), as well as their combination are applied to a linear-system solver in the shielding current analysis of a cracked high-temperature superconducting film. Although the combination seems to be the most effective of three types of the acceleration techniques, the results of computations show that, from the standpoint of the acceleration performance, neither the H-matrix-based VP nor the combination is superior to H-matrix arithmetics. The reason for this unexpected result is explained from the standpoint of operation counts.

Large AC current supply is essential for measuring the current transport characteristics of superconducting wires, such as AC loss in helical coil conductors. Commercial current supplies are generally too large and heavy, and therefore, they are inconvenient to use with the superconducting wires. The authors have been studying a small and lightweight large current supply with a high temperature superconducting (HTS) current transformer. An air-core HTS current transformer is suitable for the miniaturization of AC supply because this transformer does not have an iron core and thus does not have large volume and heavy weight. In this paper, we report that the current supply with the air-core HTS current transformer can output large current with high frequency by using the resonance phenomenon. Furthermore, we provide a method for the detection of normal transitions in the transformer by measuring the current transformation ratio of its primary and secondary currents. This detection method is composed of a simple measurement system, and experimental results show the usefulness of this detection method for the air-core transformer.

Technical verification has been progressing for high efficiency data replication between ITER and the Remote Experimentation Centre (REC) in Japan. Transferring a huge amount of data simultaneously to multiple destinations might cause excessive loads and network bandwidth on the sender so that daisy-chained relay transfer would be a considerable solution. This study demonstrates how efficiently the replication relay could be realized for the next-generation fusion experiments, such as ITER and JT-60SA. All the LHD data were consecutively sent to the REC through the global loop path (GLP: Toki - Gifu - Tokyo - Amsterdam - New York - Los Angeles - Tokyo - Aomori - Rokkasho) on SINET5 L2VPN, whose round-trip time is almost 400 ms. MMCFTP was used for the data transferring application. In both the Japan domestic path and the GLP cases, every transfer shows a very stable flattop speed as the preset 8 Gbps. However, longer gap times were needed in MMCFTP initial negotiation to establish numerous sessions. The performance optimized NVMe and iSCSI striped storages have shown higher throughputs than the ITER estimated initial data rate of 2 GB/s. Those knowledge enable the design optimization of not only the sender/receiver servers with their storages but also the intermediate relay server system.

The computer tomography for divertor impurity monitor, which measures plasma emissions in the divertor region, for ITER has been conducted using a ray-tracing technique. We have attempted four different solution methods for the inversion problem and compared the results. The solution methods which minimize errors in logarithmic scale had better performance than the methods which minimize errors in linear scale. This is likely due to the fact that the values in the emission profile vary in a wide range of orders of magnitude. The accuracy of the reconstruction has been investigated by changing discharge conditions and the number of field-of-views used. The deterioration in accuracy was most noticeable when the emission profile was reconstructed using only two field-of-views. In addition, the accuracy deteriorated, making the estimation more challenging, under discharge conditions with low emission intensity because of the wider range of emission intensity under such conditions.

Liquid lithium (Li) is one of the tritium breeders of fusion reactors and the target material of advanced fusion neutron source. However, its chemical compatibility with structural materials is one of the important issues. Austenitic steels corrode due to the preferential depletion of Ni from their surface in liquid Li. However, the mass transfer behaviors of Ni in liquid Li at low temperature have been rarely studied. The purpose of the present study is to make clear the mass transfer behaviors of Ni in liquid Li by quartz crystal microbalance (QCM). The Ni electrode of the QCM unit got wet with liquid Li of approximately 5.89 × 10⁻² cc at 473 K and 523 K for 600 seconds. The resonance frequency of the QCM unit changed due to the mass loss of the Ni electrode by the Ni dissolution into static liquid Li. The mass losses of Ni in liquid Li at 473 K and 523 K were obtained by Saurbrey's equation as 6.60 × 10⁻³ g/m² and 1.27 × 10⁻² g/m², respectively. The diffusion coefficient of Ni in static Li at 473 K was in the range between 1.13 × 10⁻¹⁰ and 5.10 × 10⁻¹⁰ m²/s by the model evaluation based on the Fick's second law. The diffusion coefficient at 523 K was in the range between 1.90 × 10⁻¹⁰ and 6.87 × 10⁻¹⁰ m²/s.

Lithium titanate (Li₂TiO₃) is candidate tritium breeder of the fusion blanket systems. The small pebbles of Li₂TiO₃ are packed in the blanket box. The coolant tubes made of RAFM steel F82H (Fe-8Cr-2W-0.1C) are installed in the blanket box and their oscillation is possibly induced by the coolant flow. The fretting corrosion is then caused by the reciprocating slip of the pebbles on the tube wall. The purpose of the present study is to investigate the fundamental behaviors of fretting corrosion between Li₂TiO₃ pebble and the F82H plate. The small pebble of Li₂TiO₃ or Al₂O₃ was placed on the plate specimen of F82H in point contact, and the plate specimen was oscillated horizontally and lineally at the frequency of 50 Hz and the amplitude of 120 µm. The constant load of 4.9 N was applied between the small pebble and the plate specimen. The fretting tests were conducted for 10 and 300 minutes in an air atmosphere at a room temperature. The Li₂TiO₃ pebble was significantly abraded in the tests. The surface of the F82H specimen was damaged due to the scratch with the broken particles of the pebble. The fretting corrosion of F82H was promoted in the test with Al₂O₃ pebble, since the formation and destruction of the oxide layer were repeated on the steel surface in the fretting cycle.

The aim of this study is to evaluate quantitatively the effectiveness of the ion exchange resins and activated carbon in tritium measurement. A standard water sample was adjusted to a tritium concentration of 5 Bq L⁻¹. This sample was stirred with the cation exchange (PCH) and anion exchange (PAO) resins and activated carbon. After stirring the samples, the supernatant in each sample was filtered through 0.45 µm membrane filters. The electrical conductivity in an experimental condition, where the amounts of PCH, PAO, and activated carbon were 0.3 g, 0.3 g, and 0.6 g, respectively, was lower than that of the sample treated by the distillation method. The distillation method requires approximately 12 h to process; however, the same level of results was achieved in 5 min. The tritium concentration in each experiment was within the standard deviation. These results suggested that proposed batch method involving ion exchange resins and activated carbon had negligible effect on tritium measurement. The ion concentration and the total organic carbon for the samples were decreased in 5 min. These results demonstrate the effectiveness of the impurity removal from the sample water by ion exchange resins.

ITER is in the integration phase and a fusion DEMO is being designed conceptually. The size of the DEMO will be around 1.5 time larger than the ITER and the weight will be about three times larger than the ITER. The TF coil and the vacuum vessel of the DEMO will become very heavy and extremely large. Although the central support structure will be used in the ITER for the assembly of TF coils and vacuum vessels, the central cylinder concept is proposed in this study to avoid the huge concentrated load at the center of the device floor during assembly. In addition, the welding joint between the vacuum vessels is discussed under the limited conditions supposed on the DEMO. Since the vacuum vessel will become the nuclear boundary, all weld lines will be expected to be examined by ultrasonic testing and/or radiographic testing to ensure the soundness of the welds.

In a three-phase matrix converter based on space-vector modulation (SVM), nine switches are controlled so that the instantaneous space vector of the line-to-line voltage rotates smoothly in two-dimensional space. The quaternion is a four-dimensional hypercomplex number that is good at describing three-dimensional rotation, such as that seen in three-dimensional game graphics programming theory. Utilizing the quaternion capability, we analyze a matrix converter by three-dimensional rotation instead of transforming to two-dimensional rotation in alpha-beta coordinates. It was clarified that the projection of the quaternion locus in three-dimensional space in the (1,1,1) direction is the same as an alpha-beta transformation locus in two-dimensional space. Concerning the direct matrix converter, we clarified that the (1,1,1)-directional superposition of three-fold higher harmonics cannot be eliminated. The quaternion can rotate and divide a three-dimensional vector. When the output voltage quaternion is divided by input one, the switching quaternion is obtained. The quaternion characteristics will be utilized to analyze a matrix converter based on direct SVM in more detail.

Liquid Sn is one of the promising coolants for liquid surface divertor concept of fusion reactors. However, the compatibility between liquid Sn and structural materials is an important issue that has to be addressed, because liquid Sn is extremely corrosive to steels at high temperatures. The corrosion may be mitigated when a protective Al₂O₃ layer is formed on the surface of alumina forming steels. However, the effect of neutron irradiation on the integrity of protective layer is not made clear so far. Japan and US joint research project “FRONTIER” started in 2019 to investigate the material compatibility under neutron irradiation. The purpose of the present study is to develop the conceptual design of the irradiation test capsule which enables material compatibility tests for the alumina forming steels - liquid metal systems under neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory, TN, USA. The three dimensional drawing of capsule structure was then developed. The validity of the material selections for the capsule design was investigated by means of corrosion tests of SiC, Si₃N₄, Ti, and Mo in liquid Sn at 773 K for 262 hr.

Stacked single-layer (SSL) coils wound with Bi2223 HTS tapes were modeled with a lumped circuit. Using the model, we investigated the effect of the joint resistance at coil terminals on the generated voltage for the SSL coils. The investigation shows that the joint resistance must be reduced as much as possible to restrain the voltage occurred in the SSL coils and each joint resistance should be uniform to make the current distribution uniform in the SSL coils. Additionally, we investigated the effect of inductance on the current distribution in the SSL coils and the effect of the tape's degradation on the coil performance. As a result, the current distribution of the SSL coils can be uniform due to the joint resistance even though the SSL coils have different inductances in each coil winding. Further, the current distribution and the voltage of the SSL coils are influenced by the tape's degradation.

Oxide Dispersion Strengthened Cu alloy (ODS-Cu) with Y₂O₃ has great application potential in the field of fusion. In the previous fabrication method by adding metal Y as the source of Y₂O₃, severe sticking occurred during mechanical alloying (MA) and huge Y particles still remained after MA. Considering that Cu-Y compounds are more brittle than pure Y, which is expected to resolve the sticking issue, and have lower Y enrichment, which will make it easier to form uniform Y distribution, ODS-Cu with various content of Y by adding Cu₂Y or Cu₆Y were fabricated through MA and Hot isostatic pressing (HIP) process. A comparative analysis was made for the samples with addition of Cu₂Y and Cu₆Y. The results showed that, compared with Cu₂Y, Cu₆Y is easier to form uniform Y distribution. The most likely reason is that the Cu₆Y is more brittle and less Y abundance than Cu₂Y. The sample with 0.39 wt% Y with Cu₆Y addition has the highest Vickers hardness, possibly because of solution strengthening caused by higher content of interstitial O, and better electrical conductivity than the sample with 1.19 wt% Y with Cu₆Y possibly because of the more uniform formation of Y₂O₃ and the absence of precipitation phases.

To study the effect of periodic plasma irradiation on deuterium (D) retention in tungsten (W), an irradiation experiment with low-energy (50 eV) and high-flux (∼10²² m⁻² s⁻¹) ions was simulated through the coupling with collision and diffusion processes of injected D atoms. The irradiation was interrupted over one and two time intervals until the total fluence reached 10²⁶ Dm⁻², in order to compare with the continuous irradiation case. The time evolution of the D atoms retained in W was calculated using the observed time variation of the material temperature and the irradiation sequences used in the experiments. After the periodic irradiation, the thermal desorption spectroscopy spectra were calculated, and the retention properties (depths, D/W ratios, and detrapping energies of the trapping sites) were estimated through comparison with the experimental spectra. At a temperature of 573 K, the irradiation with one interruption interval was found to increase the depth of the trapped D atoms and reduce the D/W ratio of the trapping sites with a low detrapping energy, compared with the continuous irradiation case. However, at 643 K, the depth was found to be decreased, and the D/W ratio of the trapping sites with a high detrapping energy was slightly increased; thus, the total retention was also slightly increased.

To detect ammonia and deuterated hydrocarbons in exhaust gas from the Large Helical Device (LHD), infrared absorption spectrometry, FTIR with a long optical path gas cell, was applied. Ammonia (NH₃) and deuterated hydrocarbons (C_xH_yD_z) could be detected during the first operations of wall baking at 368 K and the D₂ glow discharge conducted after vacuum vessel closure. The concentration of ammonia increased with increasing baking temperature, and deuterated ammonia was not detected. Thus, the ammonia, which likely originated from sweat of workers produced during vacuum vessel maintenance activities, was released from the vacuum vessel wall. Hydrocarbons were likely produced by chemical sputtering of carbon tiles and were deuterated by a hydrogen isotope exchange reaction due to D₂ glow discharge, while H₂O was released from the vacuum vessel during wall baking. It was confirmed that ammonia and various types of deuterated hydrocarbons could be measured discriminately by an FTIR spectroscopy system using a long optical path gas cell.

It is known that emission spectrum of fusion products is anisotropically distorted from Gaussian distribution when fuel ion velocity distribution function is anisotropic non-Maxwellian distribution. Previously, anisotropic neutron emission has been measured in experiments with the Large Helical Device (LHD). In this study, as an application, a method for evaluating anisotropic neutron emission spectra by measuring the spatial profiles of the neutron flux is presented. Assuming beam-heated deuterium plasma in a JT-60SA class tokamak device, it was confirmed that the effect of neutron flux emission appears in the neutron flux spatial profile outside the vacuum vessel when the energy region to be measured is set to the high energy side. In addition, it was shown that the characteristic change of the anisotropic emission spectrum is reflected in the spatial profile of the neutron flux. Estimating the neutron anisotropic emission spectrum by this measurement method can lead to new proposals for fast ion diagnosis such as analysis of velocity distribution functions.