記述言語：英語  

記述言語：英語  

QUEST計画の大きな目標の一つは、高温プラズマの定常化にかかわる研究である。本論文は、論文は論文は発表時点での長時間維持についての報告（国際球状トカマクワークショップ：平成23年11月於核融合科学研究所）をまとめたものである。特に長時間運転での粒子バランスについてのポイントモデルを用いて、QUESTのパラメータを用いると粒子バランスの定常化に必要な時間が1018秒程度になることを示した。

記述言語：英語  

After intensive discharge cleaning of the chamber wall, non-inductive current start-up experiments have been successfully performed in QUEST in moderate vertical fields of about 1.0-1.5mT with positive n-index. Simultaneously, with increasing plasma current, an asymmetric toroidal flow of energetic electrons was observed and direct measurements of current driven by this asymmetric flow were taken with a newly developed Langmuir probe technique. A numerical study of the energetic electron orbits indicates that the total current is enough to play a dominant role in the formation of a closed flux surface in QUEST. (C) 2012 American Institute of Physics. [http://dx.doi.org/10.1063/1.4731700]

DOI： 10.1063/1.4731700

記述言語：英語  

Metallic dust behavior was observed in full lower hybrid current drive plasmas on TRIAM-1M, which had all-metal plasma facing components. TRIAM-1M was shut-down on December 2005 and then in-vessel dusts were collected with a suction aspirator through a membrane filter of 100 nm meshes. The toroidal distribution of the total mass of collected in-vessel dusts was measured and the largest amount was found to be near the vertical movable limiter (VML). The typical size of in-vessel dusts was 1-10 mu m and the impact on augmenting the surface area was not significant. Dusts were observed to be either flaky or spherical in form and the majority of the flakes had accumulated around the VML Eruption detritus were observed on the surface of large flakes and this suggests that the burst of blistering are one of the origins of in-vessel dusts. (C) 2010 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2010.11.017

記述言語：英語  

Both start-up and sustainment of plasma were successfully achieved by fully non-inductive current drive using microwave with a frequency of 8.2 GHz. Plasmas current of 15 kA was implemented for 1 s. Magnetic surface reconstruction exhibited a plasma shape with an aspect ratio of below 1.5. The plasma current was dependent significantly on the launched microwave power and vertical magnetic field, while not affected by the mode of launched wave and the toroidal refractive index. Hard X-ray (HXR) emitted from energetic electrons accelerated by the microwave was observed, and the discharge with a plasma current over 4 kA followed the same trend as the number of photons of 10 keV to 12 keV. This suggests that the plasma current may be driven by energetic electrons. Based on the experimental conditions, alternative explanations of how the plasma current could be driven are discussed.

DOI： 10.1088/1009-0630/13/3/08

記述言語：英語  

A new technique to measure the current profile in plasmas with asymmetric distribution function such as lower hybrid current drive (LHCD) by using hard X-ray (HXR) energy spectrum measurement along the top and bottom identical line of sights (ILOS) is proposed and is applied to the full and partially LHCD plasmas on the TRIAM-1M tokamak at the first time in the world. The pitch angles were measured at R - R0 = (plus/minus) 2.5 cm, where R, R0 mean the major radii of the ILOS and the magnetic axis, respectively. The measured pitch angle of the magnetic field inverted at the magnetic axis estimated magnetic measurement in partially LHCD plasmas. This indicates that the difference of the measured pitch angles is caused by the plasma current in the plasma and this new method is available in detecting the current profile in tokamaks. In full LHCD plasma, no difference between the HXR signals along the top and bottom ILOS appear. This indicates that the current density around the magnetic axis was reduced compared with that in partially LHCD plasmas. This observation is no contradiction with power deposition of LHCD.

記述言語：英語  

A discharge longer than 5h was successfully achieved on TRIAM-1M by fully non-inductive lower hybrid current drive (LHCD). The heat load distribution into the plasma facing components (PFCs) during the 5 h discharge was investigated using calorimetric measurements, which estimated that the injected radio frequency (RF) power coincided with the total heat load amount to the PFCs. The power balance, including the portion of direct loss power of the fast electrons and the heat flux due to the charge exchange (CX) process, was also investigated. (c) 2006 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2006.03.004

記述言語：英語  

The first successful high power heating of a high dielectric constant spherical tokamak plasma by an electron Bernstein wave (EBW) is reported. An EBW was excited by mode conversion (MC) of an X mode cyclotron wave injected from the low magnetic field side of the TST-2 spherical tokamak. Evidence of electron heating was observed as increases in the stored energy and soft x-ray emission. The increased emission was concentrated in the plasma core region. A heating efficiency of over 50% was achieved, when the density gradient in the MC region was sufficiently steep.

DOI： 10.1103/PhysRevLett.96.185003

出版者・発行元：Nuclear Materials and Energy  

We developed and performed a 3D Monte-Carlo simulation on local erosion and deposition in outer divertor target of JA-DEMO. The local erosion and deposition were evaluated considering the divertor plasma parameters and the roof shape of W-monoblocks. First, for local erosion simulation, the trajectories of D ions and impurity (He and Ar) ions were traced before irradiating the monoblock surface. The physical sputtering was calculated by these ions irradiation, considering their charge states, incident-energies and -angles. The highest sputtering flux was produced by multiply-charged Ar ions (Ar3+–Ar5+). The sputtering flux strongly depended on the poloidal positions along the divertor target. Although almost no erosion (sputtering) was obtained in low-temperature zone (∼1 eV), the erosion occurred in mid- (∼10 eV) and high-temperature (>10 eV) zone. Second, for local deposition simulation, the trajectories of W atoms emitted from the monoblock surface were traced. After transport (including ionization and recombination) in the plasma, W atoms and ions were deposited on the monoblock surface. The deposition occurred even on the area where plasmas did not irradiate (shadow). This was due to perpendicular diffusion to magnetic field lines, gyro-orbit effects for W ions and straight-line motion for neutral W atoms. The poloidal distribution of the deposition fluxes was similar to that of the sputtering fluxes. In low-temperature zone, no deposition flux occurred due to very low sputtering flux. The deposition flux in mid-temperature zone increased significantly and that in high-temperature zone remained high.

DOI： 10.1016/j.nme.2024.101793

Web of Science

Scopus

出版者・発行元：Plasma Physics and Controlled Fusion  

A comparative investigation was conducted to assess the impact of high-field side (HFS) extraordinary (X) mode (HFS X) injection and low-field side (LFS) ordinary (O) mode (LFS O) injection of 8.2 GHz radio frequency (RF) power in the Q-shu University experimental steady-state spherical tokamak. In the case of the HFS X injection, the ratio of electron plasma frequency and RF frequency f pe / f RF was 1.3, indicating an overdense plasma, whereas in the LFS O injection, this ratio was 0.9. Distinctive features emerged in the electron temperature and density profiles between the two injection scenarios. In the HFS X injection, the profiles peaked between the fundamental electron cyclotron resonance (ECR) layer and the upper hybrid resonance layer. Conversely, in the LFS O injection, the peaks were situated near the 1st ECR layer. This implies effective excitation of the electron Bernstein wave (EBW) in the case of the HFS X injection, with the wave power being absorbed through the doppler-shifted ECR of the EBW. Density and temperature oscillations were observed only in the start-up phase of the HFS X injection, which may correlate with the presence of the left-hand cutoff of the X-mode. These findings will be contributed to understand the distinctive characteristics associated with plasma heating through EBW excitation in the HFS X injection.

DOI： 10.1088/1361-6587/ad802e

Web of Science

Scopus

出版者・発行元：Nuclear Fusion  

Superconducting (SC) tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing the ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the short circuit incident at the terminal joints of equilibrium field coil #1 during the integrated commissioning (IC) in March 2021, both EU and JA implementing agencies (IAs) have examined how to ensure safe operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum conditions are detected. Thanks to the considerable efforts of the Integrated Project Team members, the IC restarted in May 2023. After confirmation of the SC state of the coil systems (TF, EF and CS), the coil energization test and the plasma operation phase 1 (OP-1) started. The first plasma was successfully achieved on 23 October 2023 with a limited value of voltage and current applied to the coils. The plasma configuration control was also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO-F4E-QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, maintenance & enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating and diagnostic systems are extensively upgraded to allow a high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, a newly organized JT-60SA experiment team will refine the research plan for the future high heating power operation phase.

DOI： 10.1088/1741-4326/ad34e4

Web of Science

Scopus

出版者・発行元：Nuclear Fusion  

Recently, experiments on basic plasma physics issues for solving future problems in fusion energy have been performed on a Large Helical Device. There are several problems to be solved in future devices for fusion energy. Emerging issues in burning plasma are: alpha-channeling (ion heating by alpha particles), turbulence and transport in electron dominant heating helium ash exhaust, reduction of the divertor heat load. To solve these problems, understanding the basic plasma physics of (1) wave-particle interaction through (inverse) Landau damping, (2) characteristics of electron-scale (high-k) turbulence, (3) ion mixing and the isotope effect, and (4) turbulence spreading and detachment, is necessary. This overview discusses the experimental studies on these issues and turbulent transport in multi-ion plasma and other issues in the appendix.

DOI： 10.1088/1741-4326/ad3a7a

Web of Science

Scopus

出版者・発行元：Nuclear Fusion  

The plasma current start-up experiment is conducted through electron cyclotron (EC) heating in the QUEST spherical tokamak. During the EC heating, the application of a toroidal electric field in the opposite direction to the plasma current effectively inhibits the growth of energetic electrons. Observations show rapid increases in plasma current and hard x-ray count immediately following the cancellation of the retarding electric field. When a compact tokamak configuration maintains equilibrium on the high field side, along with the retarding field, it leads to effective bulk electron heating. This heating achieved an electron temperature of Te ≈ 1 keV at electron density ne > 1.0 × 1018 m−3. Ray tracing of the EC wave verifies that more power absorption into plasma through a single-pass occurs around the second resonance layer with higher values of electron density and temperature.

DOI： 10.1088/1741-4326/ad6914

Web of Science

Scopus

出版者・発行元：Nuclear Fusion  

The elimination of the need for an Ohmic heating solenoid may be the most impactful design driver for the realization of economical compact fusion tokamak reactor systems. However, this would require fully non-inductive start-up and current ramp-up from zero plasma current and low electron temperature of sub-keV to the full plasma current of ∼10-15 MA at 20-30 keV electron temperature. To address this challenge, an efficient solenoid-free start-up and ramp-up scenario utilizing a low-field-side-launched extraordinary mode at the fundamental electron cyclotron harmonic frequency (X-I) is proposed, which has more than two orders of magnitude higher electron cyclotron current drive (ECCD) efficiency than the conventional ECCD for the sub-keV start-up regime. A time dependent model was developed to simulate the start-up scenarios. For the Spherical Tokamak Advanced Reactor (STAR) (Menard et al 2023 Next-Step Low-Aspect-Ratio Tokamak Design Studies (IAEA)), it was found that to fully non-inductively ramp-up to 15 MA, it would take about 25 MW of EC power at 170 GHz. Because of the relatively large plasma volume of STAR, radiation losses must be considered. It is important to make sure that high Z impurities are kept sufficiently low during the early current start-up phase where the temperature is sub-keV range. Since the initial current ramp up takes place at a factor of ten lower density compared to the sustained regimes, it is important to transition into a higher bootstrap fraction discharge at lower density to minimize the ECCD power requirement during the densification. For the sustainment phase an array of eight gyrotron launchers with a total of about 60 MW of fundamental O-mode was found to be sufficient to provide the required axis-peaked external current drive. High efficiencies between 19-57 kA MW−1 were found with optimal aiming, and these were resilient to small changes in aiming angles and density and temperature profiles.

DOI： 10.1088/1741-4326/ad556f

Web of Science

Scopus

出版者・発行元：Nuclear Fusion  

Electron cyclotron wall conditioning with neon gas (Ne-ECWC) has been performed on the normal conducting spherial tokamak QUEST with metal walls under a trapped particle configuration with O-mode EC waves including X-mode polarization with a frequency of 8.2 GHz and an injection power of 16 kW. The Ne-ECWC removes hydrogen from the wall with small neon retention. The Ne-ECWC decreases hydrogen recycling at the following tokamak discharges, contributing to an improvement of the following tokamak plasma start-up: the plasma current increases and the start-up timing of the plasma current shifts forward. However, defects such as voids and bubbles are formed on tungsten surface exposed to the Ne-ECWC plasma.

DOI： 10.1088/1741-4326/ad3d6e

Web of Science

Scopus

記述言語：英語   出版者・発行元：九州大学グリーンテクノロジー研究教育センター  

During the early phase of tokamak plasma start-up prior to the formation of a closed magnetic surface, various plasma current profiles are expected to be observed. Especially, the noninductive current drive (CD) by the electron Bernstein wave (EBW) is likely to be the most dominant, and the current profile is expected to focus on a local area predicted by the theoretical model. In addition, the EBWCD direction could be affected by a horizontal magnetic field induced by a toroidal magnetic field coil and feed-through. The magnetic flux data cannot be fitted by using a previous model described by Yoshinaga23), and hence, this model needs improvement. Consequently, an improved model is introduced in this paper based on the theoretical prediction of EBWCD. This model can be applied to reproduce the complex plasma current profile accurately during the early phase of the tokamak plasma start-up.

DOI： 10.5109/7183306

Scopus

CiNii Research

出版者・発行元：Plasma Physics and Controlled Fusion  

The electron temperature and density profiles in the midplane of a spherical tokamak plasma produced by electron cyclotron heating (ECH) in Q-shu University experiment with steady-state spherical tokamak (QUEST) are measured by the helium line intensity ratio method. The measured profiles are compared with those obtained by the Thomson scattering method, and the measured temperatures and densities are found to agree within factors of ∼2 and ∼6, respectively. Taken together with the previous results of comparisons performed in the scrape-off layers of several toroidal devices, the same degree of agreement between the helium line intensity ratio method and other methods is obtained in the ranges of 7-100 eV for temperature and 4 × 1016-1 × 1019 m−3 for density.

DOI： 10.1088/1361-6587/ad2c2a

Web of Science

Scopus

記述言語：日本語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Nuclear Fusion  

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international Coordination on International Challenges on Long duration OPeration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed.

DOI： 10.1088/1741-4326/ad0606

Web of Science

Scopus

記述言語：日本語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Nuclear Fusion  

Results from the successful solenoid-free plasma startup using the method of transient coaxial helicity injection (transient CHI) in the QUEST spherical tokamak (ST) are reported. Unlike previous applications of CHI on HIT-II and on NSTX which required two toroidal insulating breaks to the vacuum vessel, QUEST uses a first of its kind, floating single biased electrode configuration, which does not use such a vacuum break. Instead, the CHI electrode is simply insulated from the outer lower divertor plate support structure. This configuration is much more suitable for implementation in a fusion reactor than the previous configurations. Transient CHI generated toroidal currents of 135 kA were obtained. The toroidal current during the formation of a closed flux configuration was over 50 kA. These results bode well for the application of transient CHI in a new generation of compact high-field STs and tokamaks in which the space for the central solenoid is very restricted.

DOI： 10.1088/1741-4326/ad0dd6

Web of Science

Scopus

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1007/s10894-023-00388-2

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Nuclear Fusion  

Divertor systems of fusion devices are exposed to intense heat loads from plasmas, which degrade solid plasma-facing components. Fast liquid metal (LM) flow divertors may be more advantageous for this purpose but have risk of piling due to intense magnetohydrodynamic (MHD) drag. However, severe deceleration of the flow could be countered with the injection of currents that are transverse to external magnetic fields, allowing to thrust the flow with j × B (Lorentz) forces. Given that the injection of currents as an approach to propel LM-divertor flows has remained experimentally understudied, this article focuses on the evaluation of j × B -thrust and finding its drawbacks. j × B -thrust was experimentally tested with free-surface-LM flows, a vertical magnetic field and an externally applied current. Experiments were reviewed with a theoretical model, showing agreement in the trends of theory and experiments. Full 3D-MHD-free-surface-flow simulations were also performed with FreeMHD and confirmed the sensitivity to unstable flow behavior in LM systems when applying external currents. Furthermore, excessive power requirements are expected for the implementation of j × B -thrust at the reactor scale, making these systems inefficient for commercial devices. This paper evidences that the simple operation of a LM-flow divertor with j × B -thrust, without any of the instabilities caused from reactor plasmas or parasitic currents, already presents intrinsic challenges.

DOI： 10.1088/1741-4326/ace9e9

Web of Science

Scopus

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：一般社団法人 プラズマ・核融合学会  

Liquid metal divertors are expected to have higher thermal load tolerance compared with solid divertors. Precise prediction of liquid metal behavior using magnetohydrodynamic (MHD) simulation enables the practical designing of a liquid metal divertor. The aim of this study is to confirm the accuracy of an MHD simulation that is developed for liquid metal flow prediction. As a benchmark test of the MHD simulation, a comparison of the Galinstan flowing through a vertical magnetic field and the simulated results was performed, and qualitative reproduction using the MHD simulation was confirmed. In addition, the source of wave creation and the causes of failure of the quantitative reproduction were investigated, which is may caused by the variations of the Galinstan surface tension.

DOI： 10.1585/pfr.18.2405083

Web of Science

Scopus

CiNii Research

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Nuclear Fusion  

A novel divertor biasing by four biasing plates distributed toroidally (TDDB) on the upper divertor target plate is applied to low density tokamak plasmas started-up by 28 GHz 2nd harmonic electron cyclotron current drive (ECCD) in a quasi-double null configuration of the QUEST spherical tokamak (ST). In the ST plasmas of line averaged electron density < n e > ∼0.7-1 × 1018 m−3, about 20%-40% of the current I bias driven by ∼85 V sawtooth bias voltage reaches the lower divertor plate along the biased scrape-off-layer (SOL) flux tube as the SOL current I SOL . The fact I SOL is noticeably lower than I bias indicates an appreciable leakage of parallel current from the biased SOL flux tube. The leakage currents in the toroidal and radial directions are confirmed by detection of them using the unbiased plates in the TDDB experiments. From the ion saturation current density profile obtained by a divertor Langmuir probe array, the fall-off lengths of divertor particle flux are estimated together with the strike line position. Total particle flux to the upper divertor, evaluated by the integrated ion saturation current density profile is reduced by up to 45% during positive biasing of the TDDB, depending on the position of the strike line to the biased plate. In addition, the TDDB also induces a noticeable loss of fast electrons produced by ECCD, leading to an ∼2% reduction in the maximum toroidal current of the ST plasma compared to a shot without the TDDB. Reduction of the divertor particle flux and enhancement of the fast electron losses are thought to be dominantly caused by E × B drift induced by the TDDB. In the present experimental conditions, the effects of magnetic perturbations produced by the SOL currents on the fast electron losses can be neglected because of a too small SOL current.

DOI： 10.1088/1741-4326/acee13

Web of Science

Scopus

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： 10.1585/pfr.18.2402066

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Nuclear Fusion  

Improved confinement caused by Ne injection is investigated in EAST. The safety factor q profile and electron-scale turbulence are studied with a polarimeter-interferometer system and CO2 laser collective scattering system in a series of Ne injection experiments in EAST. Ne with a suitable injection rate can lead to negative magnetic shear in the plasma core, which promotes electron-scale turbulence suppression. The continuous injection of Ne under the reversed shear condition further reduces the electron-scale turbulence intensity. Core electron temperature increase and confinement improvement are observed. QuaLiKiz is applied to model the influence of an impurity and reversed q on turbulence. The combination of stabilization effects of negative magnetic shear and Ne on turbulence is much stronger than the impurity’s own effect, which could explain the experimentally observed electron temperature (T e) increase and turbulence suppression. Besides, the relationship between pedestal structure induced by Ne injection and lower hybrid wave (LHW)-driven current deposition is discussed. These results provide a new perspective to improve the confinement in the case of Ne injection, i.e. to adjust the Ne injection to modify the ideal plasma equilibrium.

DOI： 10.1088/1741-4326/acdd83

Web of Science

Scopus

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

Here, a versatile power supply system, consisting of two current sources, insulated-gate bipolar transistor stacks, and a capacitor bank, is proposed for application in various operations of the QUEST spherical tokamak. The central solenoid coil in this system can carrying a maximum current and voltage of ±8 kA and ±375 V, respectively. In plasma start-up experiments based on ohmic heating through double flux swing, a plasma current of over 100 kA is stably obtained. Moreover, the feedback control of the thyristor power supply is utilized to control the plasma current, plasma shape, and position. The oscillating central-solenoid current applies a bipolar loop voltage at constant time intervals, and the capacitor bank discharge facilitates a co-axial helicity injection as well as amplifies the toroidal current.

DOI： 10.1016/j.fusengdes.2023.113648

Web of Science

Scopus

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：一般社団法人 プラズマ・核融合学会  

A database has been developed to provide a highly efficient data analysis environment by registering various types of data obtained in the QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) experiments and providing the data to researchers through standard SQL query output. In conventional experiments, measurements are usually obtained using several instruments, and the results are often stored separately in proprietary formats. An analysis to collect distributed files and convert them into appropriate formats individually was necessary, which led to a decrease in analysis efficiency. In order to build a system to centrally store and provide data, we constructed new servers, each for registering data in a database and for extracting data from the database and visualizing the data. As a result, various real-time monitoring data can be registered in the database, and data retrieval for analysis can be performed easily with a unified user interface. The construction of this database has made it possible to easily retrieve real-time monitoring data and a wide variety of table data, enabling analysis from a new perspective, and improving analysis efficiency.

DOI： 10.1585/pfr.18.1305048

Web of Science

Scopus

CiNii Research

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Nuclear Fusion  

While flowing Liquid Metal (LM) Plasma-Facing Components (PFCs) represent a potentially transformative technology to enable long-pulse operation with high-power exhaust for fusion reactors, Magnetohydrodynamic (MHD) drag in the conducting LM will reduce the flow speed. Experiments have been completed in the linear open-channel LMX-U device [Hvasta et al 2018 Nucl. Fusion 58 01602] for validation of MHD drag calculations with either insulating or conducting walls, with codes similar to those used to design flowing LM PFCs for a Fusion Nuclear Science Facility [Kessel et al 2019 Fusion Sci. Technol. 75 886]. We observe that the average channel flow speed decreased with the use of conducting walls and the strength of the applied transverse magnetic field. The MHD drag from the retarding Lorentz force resulted in an increase of the LM depth in the channel that ‘piled up’ near the inlet, but not the outlet. As reproduced by OpenFOAM and ANSYS CFX calculations, the magnitude and characteristics of the pileup in the flow direction increased with the applied traverse magnetic field by up to 120%, as compared to the case without an applied magnetic field, corresponding to an average velocity reduction of ∼45%. Particle tracking measurements confirmed a predicted shear in the flow speed, with the surface velocity increasing by 300%, despite the 45% drop in the average bulk speed. The MHD effect makes the bulk flow laminarized but keeps surface waves aligned along the magnetic field lines due to the anisotropy of MHD drag. The 3D fringe field and high surface velocity generate ripples around the outlet region. It was also confirmed that the MHD drag strongly depends on the conductivity of the channel walls, magnetic field, and volumetric flow rate, in agreement with the simulations and a developed analytical model. These validated models are now available to begin to determine the conditions under which the ideal LM channel design of a constant flow speed and fluid depth could be attained.

DOI： 10.1088/1741-4326/acd864

Web of Science

Scopus

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：一般社団法人 プラズマ・核融合学会  

A motor-generator slowly converts low electrical power to the large rotating energy of the induction motor, which is then converted to the higher electrical power of the synchronous generator through a pulse. Such a cycle is repeated in a tokamak plasma experiment. In the transient phase at the start of the synchronous generator, high transient voltage appears in the armature windings repeatedly. Due to the repeatability, it is imperative to estimate the high voltage and control it so that the maximum voltage is kept under a tolerable value. A quaternion is a four-dimensional hypercomplex number that is good at describing three-dimensional rotation. Utilizing the quaternion capability, a three-phase motor-generator is analyzed using three-dimensional rotation. The mechanical rotation was anlyzed by the rotation quaternion. The salient pole-type rotating field can be manipulated by direct-quadrature conversion even in quaternion analysis. The rotating dynamics and electrical phenomena of a motor-generator can be analyzed by considering the quaternion power from the motor-generator and the electrical load for plasma control.

DOI： 10.1585/pfr.18.2405035

Web of Science

Scopus

CiNii Research

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1088/2058-6272/acafc2

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://iopscience.iop.org/article/10.1088/1741-4326/acb726

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1088/1741-4326/acd4d1

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1016/j.fusengdes.2023.113794

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： DOI 10.1088/1361-6587/accaa2

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1016/j.fusengdes.2023.113462

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1016/j.fusengdes.2023.113479

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：一般社団法人 プラズマ・核融合学会  

The Thomson scattering control system has been modified to measure the time evolution of the electron density and temperature profiles during long duration discharges in the QUEST spherical tokamak. The system consists of a signal generator and a control circuit. The former accepts a QUEST main trigger and provides multiple triggers, each of which starts a short-term (e.g. 15 s) measurement. The latter provides triggers to synchronize the oscilloscope and laser oscillation during a short-term measurement. The system was used in 1000 s long duration discharges in QUEST, and the temporal evolutions of density and temperature profiles were obtained successfully. It was found the profiles are stationary after about 300 s.

DOI： 10.1585/pfr.18.1405012

Web of Science

Scopus

CiNii Research

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1063/5.0101770

記述言語：英語   出版者・発行元：九州大学大学院総合理工学府  

A new relativistic guiding center orbit code (RGCO) has been developed and used to investigate the confinement mechanism of energetic electrons in QUEST tokamak. Benefiting from the orbit simulation in cylindrical coordinates, the particle tracking can be performed within the entire first wall, including the main plasma and the scrape-off layer (SOL), so it can provide a more realistic evaluation of the confinement and loss of the particle than previous results. Detailed analyses reveal that energetic electrons can be confined both inside and outside the last closed flux surface (LCFS) inside the first wall of QUEST tokamak, thus contributing to the non-inductive current driving. In future work, the confinement and loss mechanism of alpha particle will be researched through energetic electron in QUEST by physics experiment combined with the RGCO code simulation.

DOI： 10.5109/5909096

Scopus

CiNii Research

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1063/5.0101632

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1109/TPS.2022.3193069

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1007/s10894-022-00338-4

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1016/j.fusengdes.2022.113199

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1103/PhysRevE.106.L023201

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1585/pfr.17.2405025

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1088/1741-4326/ac10e7

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Nuclear Fusion  

Spherical tokamak (ST) research in Japan has produced many innovative results: (i) plasma start-up to I p > 70 kA was achieved by electron cyclotron wave (ECW) with N ∥ = 0.75, while electron heating to T e > 500 eV was achieved with N ∥ = 0.26 on QUEST. (ii) The radiofrequency (RF)-induced transport model was combined with the x-ray emission model, and extended magnetohydrodynamics equilibrium with kinetic electrons was developed to interpret fast-electron-dominated lower hybrid wave sustained plasmas on TST-2. (iii) Density as high as 30 times the cutoff density was achieved by electron Berstein wave current drive combined with electron beam injection on LATE. (iv) Multiple plasmoids formed by tearing instability in the elongated current sheet were observed, and flux closure and ion heating by plasmoid-mediated fast magnetic reconnection were observed on HIST. (v) Optimization of ECW-assisted inductive start-up with a vertical field with positive decay index was performed on TST-2. (vi) Stabilization of the vertical displacement event by a set of upper and lower helical field coils was demonstrated on TOKASTAR-2. (vii) A 6 h discharge was achieved by cool-down of the center stack cover on QUEST, where the plasma duration limit was consistent with the wall saturation time estimated by modeling. (viii) Extension of ion heating by plasma merging was achieved on TS-3U, TS-4U, UTST, MAST, and ST40.

DOI： 10.1088/1741-4326/ac29cf

Web of Science

Scopus

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

The undamaged W (tungsten) and 0.3 dpa (displacement per atom) damaged W samples by 6 MeV Fe2+ (Fe ion) irradiation were installed in the first wall of QUEST (Q-shu University Experiment with Steady-State-Spherical Tokamak) device and exposed to H (hydrogen) plasma in 2018A/W (Autumn / Winter) or 2019S/S (Spring / Summer) campaign to evaluate the impact of damages and impurities on hydrogen isotope retention. The surface morphology and chemical states of constituent atoms were observed by TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). It was found that thick Al deposit were found for the samples in 2019S/S campaign, which would come from the insulating plate during the CHI (Coaxial Helicity Injection) discharge. The additional 1 keV D2+ was implanted into both of these samples and D (deuterium) retention enhancement was evaluated by TDS (Thermal Desorption Spectroscopy). The downward ion toroidal drift changed the impurity deposition and damage profiles. In 2018A/W and 2019S/S, the D retentions for undamaged W samples had position independent, indicating that plasma would be well-controlled in this configuration. In case of Fe2+ damaged samples, irradiation damages clearly changed the D retention characteristics.

DOI： 10.1016/j.fusengdes.2022.113020

Web of Science

Scopus

記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：Plasma Science and Technology  

An internal transport barrier (ITB) can be formed on EAST in exploring high-parameter operation. Previous studies show that safety factor (q) profiles, Shafranov shift and magnetohydrodynamic behaviors could be helpful in ITB formation by suppressing anomalous transport. Recently, electron density evolution with high resolution demonstrates that fishbone could be dominant in electron density ITB formation and sustainment. The power threshold is low in the fishbone condition and the electron density profile is determined by traits of fishbone. Simulation shows that the low-k ion mode is suppressed by fishbone. Direct measurement of turbulence in the inner region shows that the internal kink mode could sustain an electron temperature ITB by suppressing the trapped electron mode. The multi-scale interaction between the kink mode and turbulence by current could be key in sustaining high-electron-temperature long-pulse operation.

DOI： 10.1088/2058-6272/ac2726

Web of Science

Scopus

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： 10.1088/1361-6587/ac1838

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： 10.1088/1361-6587/ac1b1f

記述言語：英語   掲載種別：研究論文（学術雑誌）  

QUEST (Q-shu university experiment with steady state spherical tokamak) is a midsize spherical tokamak capable of steady-state operation, comprising all-metal plasma-facing walls and a hot wall (HW) to address issues pertaining to fuel particle balance. The HW was installed summer 2014. Quantitative analysis pertaining to the HW at 373 K is carried out, and clarify the quantitative impact of shot history that obviously appears in wall stored hydrogen just before the discharge at the wall temperature. The model indicates the plasma-induced deposition layer play an essential role in fuel particle balance. A clear temperature dependence of fuel recycling was observed using outgassing just after plasma termination and played an essential role in regulation of particle balance. Consequently, long duration discharges lasting more than 1 h has been obtained at wall temperature, TW < 423 K in the range of 40–60 kW of RF power and, similarly, a 6 h discharge could be achieved at TW = 473 K using 20 kW of injected RF power.

DOI： 10.1016/j.fusengdes.2021.112436

記述言語：日本語   掲載種別：研究論文（学術雑誌）  

The W (tungsten) samples were placed at Top, Equator and Bottom plasma facing walls of QUEST (Q-shu University Experiment with Steady-State-Spherical Tokamak) device and exposed to 754 shots of hydrogen plasma during 2017A/W (Autumn/Winter) campaign. Thereafter, their surface morphologies and chemical states were evaluated by TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). The XPS results showed that a thick carbon layer about 3–18 nm has formed throughout the wall surface. Among them, the Bottom wall had the deposition layer with the thickness of 2 nm, which was thinner than the top wall, namely erosion-dominated. On the other hand, a thick C layer about 18 nm was deposited on the Equator wall. The additional 1 keV D2+ was implanted into these samples and the D (deuterium) retention enhancement was estimated. The D2 TDS (Thermal Desorption Spectroscopy) spectra for all the samples had two major desorption stages at 400 K and 650 K, namely the desorption of D trapped by irradiation damages and deposition layer. The erosion/deposition profile would be caused by wall position and plasma condition, like a current start-up experiment. The desorption temperature of H2 (hydrogen) was shifted toward higher temperature side compared to that exposed to previous plasma campaign (2016 A/W), suggesting that H was mainly accumulated in the deposition layer with forming C–H bonds.

DOI： https://doi.org/10.1016/j.nme.2020.100856

記述言語：日本語   掲載種別：研究論文（学術雑誌）  

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.

DOI： https://doi.org/10.1585/pfr.16.2402048

記述言語：日本語   掲載種別：研究論文（学術雑誌）  

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.

DOI： https://doi.org/10.1585/pfr.16.2405037

記述言語：日本語   掲載種別：研究論文（学術雑誌）  

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.

DOI： https://doi.org/10.1585/pfr.16.2402024

記述言語：日本語   掲載種別：研究論文（学術雑誌）  

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.

DOI： https://doi.org/10.1585/pfr.16.2402034

記述言語：日本語   掲載種別：研究論文（学術雑誌）  

A spectroscopic system with multiple viewing chords was developed for QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) to measure the spatial distribution of ion toroidal velocities in discharges sustained by electron cyclotron resonance heating (ECH). Twenty-four viewing chords were aligned in the midplane and C III emission line spectra were measured for three types of ECH discharge under different magnetic field configurations. By applying an inversion method to the measured spectra, we evaluated the radial distributions of C2+ ion emissivity, temperature, and toroidal velocity. The error in the evaluated velocity was estimated to be less than 5 km/s. It was found that the velocity depends on the magnetic field configuration.

DOI： https://doi.org/10.1016/j.nme.2021.100905

記述言語：日本語   掲載種別：研究論文（学術雑誌）  

Abstract
To study radial distribution of hydrogen flux in peripheral plasma in long duration tokamak discharges, permeation and Langmuir reciprocate probes are used. The atomic hydrogen flux to the walls at several radial positions is deduced from the permeation flux. Contribution of the ions to the hydrogen flow is deduced from the Langmuir probe ion saturation current. It is found, that the atomic hydrogen has the most contribution to the hydrogen flow behind the plasma facing components (PFCs), while the ion flux there is negligible. The atomic flux behind the PFCs drops to the value of , which is 6% of that just in front of PFCs, .

DOI： https://doi.org/10.1016/j.nme.2020.100872

記述言語：日本語   掲載種別：研究論文（学術雑誌）  

The plasma current is ramped up primarily by a 28 GHz electron cyclotron wave (ECW) in the Q-shu University experiment Steady-State Spherical Tokamak (QUEST), with multiple harmonic resonance layers from the second to the fourth stay in the plasma core. A steering antenna comprising two quasi-optical mirrors enhances the power density of ECWs. The ECW beam is injected obliquely from the low-field side where the parallel refractive index is N∥ = 0.75 at the second-harmonic resonance layer. Analysis of the resonance condition has found that energetic electrons moving forward along the magnetic field resonate more effectively than those moving backward. Such symmetry breaking is consistent with the results of the current ramp-up experiment. The peak plasma current reaches Ip>70 kA, constantly injecting a beam of radio frequency power of 100 kW. Ray-tracing by the TASK/WR code demonstrates that the power of the 28 GHz extraordinary mode is absorbed by energetic electrons via single-pass cyclotron absorption.
I. INTRODUCTION

DOI： https://doi.org/10.1063/5.0031357

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1016/j.fusengdes.2021.112436

記述言語：英語  

DOI： 10.1002/fld.4906

記述言語：英語   掲載種別：研究論文（学術雑誌）  

DOI： https://doi.org/10.1585/pfr.15.2402063

記述言語：英語  

DOI： 10.1063/5.0013966

記述言語：英語  

DOI： 10.1088/1741-4326/aba0c6

記述言語：英語  

As technology has improved in recent years, it has become possible to create new valuable functions by combining various devices and sensors in a network. This concept is referred to as the Internet of Things (IoT), and predictive maintenance is a new valuable function associated with the IoT. In large-scale experimental facilities with many researchers, it is not desirable that experiments cannot be performed due to sudden failure of equipment. For this reason, it is important to predict the failure in advance based on the measurement results of sensors and to perform repairs in a planned manner. On the Q-shu University experiment with steady-state spherical tokamak (QUEST) large experimental device, it is necessary to drive a large current of 50 kA, and the diagnosis of its power line deterioration is well performed as predictive maintenance through the evaluation of its contact resistances of several micro ohms order on the network. In addition, as an example of the IoT, mechanisms to assist safe operation, such as a sound alarm system and an entrance management system, are built by sharing experimental information between devices via the network.

DOI： 10.1016/j.heliyon.2020.e04214

記述言語：英語  

DOI： 10.3390/atoms8030044

その他リンク： https://doi.org/10.3390/atoms8030044

記述言語：英語   掲載種別：研究論文（学術雑誌）  

Fast Ejecting System of Targeted sAmple called FESTA has been developed to carry out the measurement of dynamic hydrogen retention by a test sample. A sample can be exposed and extracted from the targeted plasma at any time using FESTA, however, when exposing the sample, the test chamber wall gets coated by some hydrogen as it is open to the QUEST vacuum vessel. We refer to this as the plasma-induced background. To measure the amount of hydrogen retained by the sample itself, the contribution from the plasma induced background must be subtracted from the measurements. To measure the accurate dynamic retention from plasma-exposed sample, a background subtraction model has been developed and tested. The initial testing shows that the FESTA system and model can estimate the dynamic hydrogen retention by a target test sample.

DOI： 10.1585/PFR.15.2402013

記述言語：英語  

As part of the US-Japan bilateral collaboration program, a new series of workshops have been inaugurated on the subject of 'Power and Particle Control by Liquid Metal Plasma-Facing Components in a Steady State Magnetic Fusion DEMO Reactor'. The first workshop of this series was held at Princeton Plasma Physics Laboratory on 11-13 March 2019. This conference report is intended to share 24 technical presentations, discussion and the planning of future workshops with the world magnetic fusion community.

DOI： 10.1088/1741-4326/ab4ede

記述言語：英語  

Figure 8 in [1] was partially incorrect because of the heat flux data derived from Langmuir probes which included the effect of arcing in long-pulse high-power heating plasmas on EAST tokamak. The new plot with corrected data and errorbar for figure 8 is presented here. (Figure Presented).

DOI： 10.1088/1361-6587/ab6b03

記述言語：英語  

A 28 GHz system with a high-power gyrotron tube has been used for the QUEST spherical tokamak to form an over-dense plasma for electron Bernstein wave (EBW) heating and current drive with an 8.2 GHz-wave. Non-inductive high-density plasma ramp-up experiments with dual-frequency (dual-f ) electron cyclotron (EC) (8.2 GHz and 28 GHz) waves were conducted. A spontaneous density jump (SDJ) to an over-dense state was first observed as a bifurcation phenomenon in the dual-f  wave experiment. The over-dense plasma on the 8.2 GHz wave was non-inductively ramped up to 25 kA, and was maintained for 0.4 s under stable plasma equilibrium after two such jumps in one shot. Heating to mildly energetic electrons and bulk electrons was observed even in the over-dense region. The electrostatic EBW heating effect on the mildly energetic electrons in the over-dense region is assessed following a dispersion analysis of the 8.2 GHz wave. The bulk electron heating effect observed is explained as heat exchange from mildly energetic electrons heated by the electrostatic EBW. Remarkably, a high hard-x-ray-radiation temperature of ∼500 keV was also observed in tangential viewing for current-carrying electrons in the over-dense core region. Synergetic heating from the overlap of different 28 GHz EC harmonic resonances as well as higher harmonic heating is discussed for maintaining the highly energetic electrons in the over-dense core region. In addition, the SDJ process and mechanism are considered based on the discussion of the electron heating effects with the 8.2 GHz wave.

DOI： 10.1088/1741-4326/ab4c12

記述言語：英語  

This paper presents a detailed design of the Q-shu University experimental steady state spherical tokamak's (QUEST's) high-field side (HFS) injection system for electron Bernstein wave (EBW) excitation and the results of an experimental comparison of the HFS eXtraordinary X-mode and low-field side (LFS) ordinary O-mode injection of 8.2 GHz radio frequency (RF) power. Waveguides, as an alternative to mirror polarizers for transmitting RF X-mode power from LFS to HFS for EBW conversion, were used instead of the installation of an RF mirror. Testing of LFS-to-HFS RF power transmission at 8.2 GHz, using an RG-50-type vacuum waveguide in a bench-scale device filled with SF₆ gas at 0.03 Mpa, revealed that an RF power of 10.8 kW could traverse the fundamental electron cyclotron resonance layer for 60 s without breakdown. The short-length, open-ended waveguide antenna used in the HFS injection-induced wave diffraction reduced the efficiency of power delivery to the upper hybrid resonance (UHR) by approximately 7&#37; at an electron temperature of 50 eV. The HFS injection was able to produce brighter camera images than the standard LFS injection. The location of the UHR, as estimated by measuring the density with an interferometer, agreed with its location as measured by plasma radiation low-field, side-edge positions shown by fast camera imaging. This indicates that the plasma was produced by mode-converted EBW. The HFS injection had an absorption efficiency of 96&#37;, compared to 40&#37; for LFS. A greater fluctuation of floating potential adjustable to the lower hybrid wave (LHW) was observed in the HFS case by installing a Langmuir probe, confirming that EBW conversion efficiency was higher in the HFS case. Moreover, after setting the poloidal field to B_PF = 7.6 mT, plasma current (I_P ) in the HFS peaked at 1.3 kA, as opposed to 0.3 kA for LFS, despite LFS injection having a total power of 55 kW, compared to 40 kW for HFS. However, as the impurity level was comparatively high, it is believed that this I_P is dominated by pressure-drive, which makes it difficult to analyze EBWCD. Finally, the line-integrated density in the HFS injection peaked at 1.6 × 10¹⁸ m^-2, compared to 8 × 10¹⁷ m^-2 in the LFS one.

DOI： 10.1088/1361-6587/ab6903

記述言語：英語  

In order to address the problem of heat and particle removal in tokamak-type, magnetic confinement nuclear fusion reactors, a divertor that utilizes liquid metal is suggested to replace solid tungsten divertors due to concerns regarding their structural integrity at high energy fluxes. The operation of such a device gives rise to phenomena spanning multiple disciplines of physics such as fluid dynamics, electromagnetics, thermodynamics and plasma physics. Smoothed Particle Hydrodynamics (SPH) is a Lagrangian, mesh-free numerical method that has been proven effective in a variety of disciplines. In this work, the hydrodynamic aspect of the liquid metal divertor is simulated using SPH, paving the way to implement additional physics in future work.

記述言語：英語  

The W (tungsten) samples were placed at top, equator and bottom walls of QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) device and exposed 1238 shots of hydrogen plasma during 2016A/W (Autumn/Winter) campaign with normal wall temperature of 473 K (maximum temperature of 523 K). Thereafter, the surface morphology was evaluated by color measurement, TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). Thick deposition layers were formed on the samples placed at the equator and bottom walls. On the other hand, thin mixed material layer was deposited on the top wall, where large H (hydrogen) retention was observed, which would be caused by dynamic plasma wall interaction (erosion and deposition) with higher H flux. Low H retention was confirmed for bottom wall, where higher wall temperature without He discharge would contribute. The additional 1 keV D₂ ⁺ was implanted into these samples and deuterium retention enhancement was estimated. It was clearly found that the irradiation damages would induce more deuterium trapping than the formation of C–D bond.

DOI： 10.1016/j.fusengdes.2019.02.110

記述言語：英語  

In the EC-driven (8.2 GHz) steady-state plasma on QUEST, plasma current seems to flow in the open magnetic surface in the outside of the closed magnetic surface in the low-field region according to plasma current fitting method without taking equilibrium into account. In our previous work, plasma equilibrium solution was fitted assuming all plasma current is flowing in the inside of the Last Closed Flux Surface (LCFS). It was solved within isotropic pressure profile by EFIT code. Opposite-polarity current density region appeared in the high-field region. Here in this article, considering the toroidal rotation, the equilibrium is fitted within nested magnetic surfaces by SU-EFIT. Though the plasma magnetic axis shifts outward due to the centrifugal force, the opposite polarity current does not disappear in the high-field region. And relation between the toroidal rotation speed and the poloidal beta value will be discussed.

DOI： 10.1016/j.fusengdes.2019.04.059

記述言語：英語  

A high voltage DC power supply for the cathode of the 8.56 GHz CW klystron has been set up for electron cyclotron heating (ECH) in steady state tokamak operation on QUEST spherical tokamak. The power supply is equipped with an IGBT array and a reactor for fast shutoff of the voltage in 10 μs, where the influx of electric energy at the short circuit is limited to 5 J. AC switches also have been installed in the three-phase power lines. High voltage relays are useful to save electric energy consumption. Fast three-phase AC switching by IGBT-stack is applicable to reduce the electric load of the components of the klystron power supply.

DOI： 10.1016/j.fusengdes.2019.04.043

記述言語：日本語  

DOI： 10.1585/pfr.14.1402128

記述言語：日本語  

DOI： 10.1585/pfr.14.1405125

記述言語：英語  

Megawatt (MW) gyrotrons, with a wide frequency range from 14 to 300 GHz, are being developed as part of a collaborative electron cyclotron heating (ECH) study for advanced fusion devices and a demonstration power plant (DEMO). (1) Detailed designs for a 14 GHz 1 MW gyrotron are being developed for fabrication. For a 14 GHz radio frequency (RF) beam with high divergence, a calculated transmission efficiency of 94&#37; to the corrugated waveguide coupling position was obtained initially by introducing the design concept (direct RF beam coupling by built-in waveguide) to minimize the RF transmission path. Installing a double-disk sapphire window will make it possible to develop a 1 MW gyrotron with a continuous wave (CW) at 14 GHz. (2) In experimental tests of a new 28/35 GHz dual-frequency gyrotron, the cooling characteristics of an optimal-structure double-disk sapphire window were evaluated. We confirmed that operating at 0.4 MW with a CW at 28 GHz is feasible, reaching twice the output power reported in previous studies. In a 2 ms short-pulse experimental test, maximum powers of 1.65 MW at 28.04 GHz and 1.21 MW at 34.83 GHz were achieved. (3) A design study of a 77/51 GHz dual-frequency gyrotron was performed. Oscillations above 1.5 MW for 77 GHz and 1.3 MW for 51.88 GHz are expected for a beam voltage V _k = 80 kV and beam current I _k = 60 A. (4) In an experiment with a 300 GHz gyrotron, the influence of the wave reflected from the window was reduced by tilting the output window, and mode competition in the cavity was suppressed. An output power of 0.62 MW with a pulse width of 1 ms, which is the new record for this frequency, was obtained. (5) We also performed a trial design study of a 240 GHz gyrotron for DEMO.

DOI： 10.1088/1741-4326/ab0e2c

記述言語：英語  

New polarizer and launcher systems on a 28-GHz electron-cyclotron (EC) heating and current drive (HCD) system have been developed for non-inductive second-harmonic EC-plasma-current ramp-up in the QUEST spherical tokamak. A launcher system with two quasi-optical mirrors providing beam steering capability was designed to focus the incident beam to a small-sized waist of ∼0.05 m at the second-harmonic EC resonance layer. A relatively large focusing mirror was designed based on a Kirchhoff integral code developed to derive wave solutions. The focusing property of the launched beam was first confirmed with a 3-dimensional (3D) electromagnetic-wave simulator. Focusing characteristics were also checked at low-power test facilities, together with the steering capability. The performance of this launcher system was demonstrated to work as designed, and assembled in the QUEST device. The system was applied to the non-inductive second-harmonic EC plasma ramp-up experiments with no optimization required regarding the incident polarization. The results obtained for the non-inductive plasma ramp-up are also presented.

DOI： 10.1016/j.fusengdes.2019.02.027

記述言語：英語  

High field side (HFS) injection of eXtra-ordinary X-mode for electron Bernstein wave (EBW) conversion was conducted in the QUEST tokamak. Radio frequency (RF; 8.2 GHz) power was delivered from the low field side (LFS) to the high field side HFS through waveguides, and from the HFS placed 20 cm above the midplane of the vacuum vessel. The aim was to compare the RF launches from the LFS and HFS. The plasma brightness, measured by a fast camera, as well as the H
_α
signal captured along the mid-plane, was noticeably higher in the HFS launch than in the LFS launch. The HFS injection achieved a plasma current of approximately 130 A, versus 35A in the LFS injection. The electron density n
_e
predicted from the position of the upper hybrid resonance agreed with the line-averaged n
_e
measured by an interferometer, confirming the effective conversion and subsequent damping of the EBW mode. The RF leakage of the HFS injection was less than one-sixth that of the LFS injection. These results indicate that HFS delivers better RF coupling and conversion efficiency to EBW than LFS injection. Such efficient plasma heating via EBW will significantly enhance the plasma production.

DOI： 10.1585/pfr.14.1205038

記述言語：英語  

Initial results from the recently implemented transient coaxial helicity injection (CHI) system on QUEST are reported. QUEST uses a new CHI electrode configuration in which the CHI insulator is not part of the vacuum boundary, making this configuration easier to implement in fusion reactors. Experimental results show that transient CHI startup in this alternate electrode configuration is indeed possible. Reliable gas breakdown was achieved, and toroidal currents up to 45 kA were generated.

DOI： 10.1088/1361-6587/aadcb7

記述言語：英語  

Dependence of heating temperature on tritium retention behavior in stainless steel type 316 L (SS316L) has been examined about each process of degassing and tritium exposure. Two kinds of SS316L samples were employed: bare SS316L and plasma-exposed SS316L. The amount of tritium retained in surface layers of a sample was nondestructively measured by β-ray-induced X-ray spectrometry, and changes in the surface state with heating in vacuum were analyzed by X-ray photoelectron spectroscopy (XPS). Significant increase in tritium retention in bare SS316L samples appeared in a degassing temperature region above 600 K. Similar tendency of tritium retention was observed for the plasma-exposed sample. It was seen that the degassing process prior to tritium exposure significantly influenced to the tritium retention behavior. Furthermore, it was suggested from surface analysis by XPS that chemical states of SS316L surface at high temperatures play an important role for tritium retention behavior.

DOI： 10.1016/j.nme.2018.05.024

記述言語：英語  

Metal plasma-facing materials (PFMs) are expected to be candidates for future fusion power plants from the view point of tritium retention. The purpose of this study is to develop a model including a long timescale plasma interaction with metal PFMs. As a first step, we have developed a simple zero-dimensional (0D) model, which consists of particle balance equations for the following three different particle species: (a) hydrogen plasma (elec., H⁺, H⁺ ₂ , H⁺ ₃), (b) neutral hydrogen atoms (H) and molecules (H₂) in the gas phase, and (c) the wall-stored H atoms. The model has been applied to simulate long-term operation in the limiter configuration of the QUEST tokamak. Modelling results of the long time evolution of the H atom wall inventory reasonably reproduce the experimental tendency. Although the present model is relatively simple, it is useful to understand the basic characteristics of overall plasma particle balance, the density control of the main plasma, and the H atom wall inventory for long-term tokamak operation.

DOI： 10.1002/ctpp.201700129

記述言語：英語  

Intensity fluctuations are investigated using the fast camera imaging technique in the slab annular plasma as a function of magnetic shear and connection length in the spherical tokamak QUEST. Note that here QUEST is operated as a simple magnetized torus with a tight aspect ratio. Slab annular plasmas feature open magnetic field lines and can mimic the tokamak edge-scrape off layer (SOL)-like plasma attributes reasonably well. Three magnetic shear regimes are realized using three poloidal magnetic field (PF) coil pairs. A whole range of connection lengths (∼∞ ≥ L _c ≥ 5.5 m) is scanned by varying the PF strength for a given toroidal field for each magnetic shear regime. For the first time a systematic study of the effect of magnetic shear and field line pitch together on edge-SOL-like plasma fluctuations is being reported. Slab plasmas with intermediate magnetic shear are observed to be more susceptible to generate distinct blobs when L _c is reduced by increasing the PF strength. A distinct coherent mode appears only at the lowest magnetic shear slab featuring a deep potential well. Such mode is not apparent at other magnetic shear cases even at the same L _c. Finally, with a combination of PF coil pairs, both the features of intermediate and low magnetic shear slabs are shown to be realizable simultaneously. Significantly stronger blobs are observed with such combination of PF mirror ratios in the presence of a coherent mode. This study may provide better insight into the effect of magnetic configuration in the tokamak edge and SOL turbulence and can help in searching for better tools to control cross-field convective intermittent transport in tokamaks.

DOI： 10.1088/1361-6587/aacb69

記述言語：英語  

In tokamaks, the temperature of the plasma-facing wall is an important parameter for achieving particle balance and therefore steady-state operation. QUEST, which is a middle-sized spherical tokamak, has hot walls that act as plasma-facing walls. They can be actively heated with sheath heaters and actively cooled with water. To control the wall temperature, heating and cooling systems have been developed. These systems adjust the power of the sheath heaters and the motor valves of the cooling system, respectively. The two systems communicate via Ethernet through UDP and control the hot-wall temperature cooperatively. The plasma control system (PCS) in QUEST has also been modified, especially with respect to gas fueling, in order to enable long-duration plasma sustainment. A feedback controller has been installed in the PCS, together with a mass flow controller, allowing Hα emission from the plasma which is used as a reference signal, to be well controlled. Plasma density calculations using a field-programmable gate array are proposed for the feedback control system.

DOI： 10.1016/j.fusengdes.2018.02.069

記述言語：英語  

An evaluation of high-field-side (HFS) X-mode injection for the electron-Bernstein-wave (EBW) scenario is performed using the GENRAY ray-tracing code. In the early stage of low-density plasma start-up, when the electron cyclotron resonance and upper hybrid resonance layers are close to each other, efficient and localized heating by the EBW is attainable. We show that, when the electron density rises, the HFS scenario spontaneously shifts to current drive with successful electron heating. This shift can be explained as a change in heating mechanism from collisional to electron cyclotron damping. Also, we discuss a possible O-X-B scenario to continue the plasma current drive beyond the formation of an over-dense plasma.

DOI： 10.1585/PFR.13.3402115

記述言語：英語  

In QUEST, transient Coaxial Helicity Injection (CHI) has now been implemented using a new electrode configuration in which the CHI insulator is not part of the vacuum boundary. In this paper, for the first time, suitable conditions for generation of the CHI-produced toroidal current in the QUEST vessel configuration were investigated using the Tokamak Simulation Code (TSC). The simulation results show that the configuration in which the biased electrode is located farther away from the injector flux coil requires higher currents in the injector coil to generate the required injector flux. Additionally, energizing a lower inboard poloidal field coil and possibly lowering the electrode plate closer to the injector flux coil may be necessary to improve injector flux shaping to permit a configuration that is more favorable for inducing flux closure.

DOI： 10.1585/PFR.13.3402059

記述言語：英語  

Toroidal electron cyclotron resonance (ECR) plasma is an ECR heated plasma in an open toroidal magnetic field. The plasma contains no toroidal current and is used for the pre-ionization and non-inductive startup of a tokamak plasma. To obtain a deeper understanding of the basic properties of the plasmas produced in the spherical tokamak QUEST (Q-shu University Experiment with steady-state Spherical Tokamak), we have developed an optical emission spectroscopy system with multiple viewing chords and have used it to measure the spatial distributions of the toroidal and poloidal flow velocities of C²⁺ ions. We compare the measured velocities with those calculated from the ion drift equations using the plasma parameters reported for a similar spherical tokamak, LATE (the Low Aspect Ratio Torus Experiment).

DOI： 10.1585/PFR.13.3402087

記述言語：英語  

In the present RF-driven (ECCD) steady-state plasma on QUEST, the plasma current seems to flow in the open magnetic surface outside of the closed magnetic surface in the low-field region according to the plasma current fitting (PCF) method. The current in the open magnetic surface is due to the orbit-driven current by high-energy particles in the RF-driven plasma. High-energy particles guiding center orbits are calculated as contour plots of conserved variables in the Hamiltonian formulation considering particles in the initial position with different energies and pitch angles satisfying the resonance condition. A negative current appears near the magnetic axis, and a hollow current profile is expected even if the pressure driven current is considered. The equilibrium is fitted within nested magnetic surfaces by J-EFIT coded by MATLAB using the hollow current profile shift toward the low-field region. Although the plasma boundary shape reflects the plasma current density profile, the equilibrium shape fitted by J-EFIT does not coincide with the orbit-driven current profile. However, introducing an extension of the current profile without assuming nested contours into the J-EFIT code appropriately fits the plasma shape with the hollow current profile to the measured magnetic data.

DOI： 10.1016/j.fusengdes.2017.05.070

記述言語：英語  

DOI： 10.1088/1741-4326/aa7c20

記述言語：英語  

Fully non-inductive second (2nd) harmonic electron cyclotron (EC) plasma current ramp-up was demonstrated with a newlly developed 28 GHz system in the QUEST spherical tokamak. A high plasma current of 54 kA was non-inductively ramped up and sustained stably for 0.9 s with a 270 kW 28 GHz wave. A higher plasma current of 66 kA was also non-inductively achieved with a slow ramp-up of the vertical field. We have achieved a significantly higher plasma current than those achieved previously with the 2nd harmonic EC waves. This fully non-inductive 2nd harmonic EC plasma ramp-up method might be useful for future burning plasma devices and fusion reactors, in particular for operations at half magnetic field with the same EC heating equipment.

DOI： 10.1088/1741-4326/aa7c20

記述言語：英語  

After several experimental campaigns in the Kyushu University Experiment with Steady-state Spherical Tokamak (QUEST), the originally stainless steel plasma-facing wall (PFW) becomes completely covered with a deposited film composed of mixture materials, such as iron, chromium, carbon, and tungsten. In this work, an innovative colorimetry-based method was developed to measure the thickness of the deposited film on the actual QUEST wall. Because the optical constants of the deposited film on the PFW were position-dependent and the extinction coefficient k₁ was about 1.0-2.0, which made the probing light not penetrate through some thick deposited films, the colorimetry method developed can only provide a rough value range of thickness of the metal-containing film deposited on the actual PFW in QUEST. However, the use of colorimetry is of great benefit to large-area inspections and to radioactive materials in future fusion devices that will be strictly prohibited from being taken out of the limited area.

DOI： 10.1063/1.5000739

記述言語：英語  

In order to understand the atomic hydrogen distribution in different kinds of plasma and its influence on the recycling, two kinds of plasmas were used: non-confined annular electron cyclotron resonance (ECR) and confined long duration plasmas. The permeation probes are used to measure directly the atomic hydrogen flux at several poloidal positions. The permeation through metals due to the ion and atom component of the hydrogen flux to the wall is indistinguishable. To estimate the contribution of the ions directly, Langmuir probes were used. The Г_inc profile behind the plasma facing components (PFCs) is almost constant, ∼2 ×10¹⁸ H/s/m².

DOI： 10.1016/j.nme.2017.03.027

記述言語：英語  

An efficient technique for building plasma equilibria with preset or constrained q profiles by solving the nonlinear Grad-Shafranov equation iteratively is proposed. The technique uses a set of dynamic virtual poloidal field probes in the plasma core region. Numerical examples are shown for monotonic q profiles and negative central magnetic shear equilibria which are obtained with this technique. As a proof of principle, the technique is applied on the EAST tokamak, to constrain the plasma equilibrium reconstruction with the experimental q = 1 surface information from the ECE diagnostic in addition to the external magnetic measurements.

DOI： 10.1088/1741-4326/aa74eb

記述言語：英語  

Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.

DOI： 10.1088/1741-4326/aa62c1

記述言語：英語  

In tokamak operations, breakdown of plasma is the first step of the plasma build-up. In this paper, we present a combinative investigation of radio frequency (RF)-induced breakdown experiments in QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) and a one-point model of hydrogen ionization. Experimental results with two different frequencies of 2.45 GHz and 8.2 GHz showed that the clear threshold on connection length, L, existed for breakdown with a negative n-index configuration n=-(R/Bv)·(∂Bv/∂R), where R is the major radius and B_v the is vertical magnetic field. In contrast, breakdown was always obtained with positive n-index when changing L. It indicates that a lifetime of an incubated electron plays a significant role in the plasma breakdown. According to one-point model calculation, the experimental threshold of L is well predicted by the lifetime of the incubated electron estimated by employing the loss term along with L. The model calculation also describes the requirement of the minimum electron temperature T_e for RF-induced breakdown to realize an avalanche of electrons in the tokamak magnetic structure.

DOI： 10.1063/1.4985142

記述言語：英語  

Megawatt power gyrotrons are being developed for collaborative electron cyclotron heating (ECH) studies of advanced fusion devices and demonstration power plant (DEMO). (1) In the first experiment of a 300 GHz gyrotron, an output power above 0.5 MW in the TE_32,18 single mode was achieved with a pulse width of 2 ms. This was the first observation of MW-scale oscillations in a DEMO-relevant gyrotron mode. It was also found that the reflection at the output window affected the determination of the oscillation mode. Furthermore, several single mode oscillations in the 226-254 GHz range were confirmed, which is important for the step-frequency tunable gyrotron in the sub-THz frequency range. (2) Based on the successful results of the 77 and 154 GHz large helical device (LHD) tubes, a new 154/116 GHz dual-frequency gyrotron with an output of over 1.5 MW is being designed. (3) A new record output of 1.38 MW was obtained using an existing 28 GHz gyrotron. A newly designed tube aimed at achieving a dual-frequency output power of 2 MW at 28 GHz (0.4 MW continuous wave) and 1 MW at 35 GHz was built. In the first experimental test, main mode oscillations were observed at the frequencies of 28.036 and 34.831 GHz with Gaussian-like output beams and output power of 1.27 and 0.48 MW, respectively. A total efficiency of 50&#37; was achieved at 28 GHz operation.

DOI： 10.1088/1741-4326/aa6875

記述言語：英語  

The first equilibrium reconstruction of EAST current-density profile based on internal Faraday rotation measurements provided by the POlarimeter-INTerferometer (POINT) diagnostic is demonstrated using the EFIT equilibrium reconstruction code. EFIT incorporates 11 simultaneous line-integrated density and Faraday effect measurements from POINT to self-consistently reconstruct the equilibrium toroidal current density profile using a Faraday rotation reconstruction algorithm. It is shown that the POINT measurements can be applied to improve the accuracy of core plasma current density and q profile on EAST. Comparisons of magnetic surfaces and the q profile reconstructed using external magnetic data against those using magnetic and POINT data are presented. Equilibrium reconstructions using POINT data are found to be consistent with sawtooth phenomena. The sensitivity of equilibrium reconstruction to POINT measurements indicates Faraday rotation provides important constraints for determining the current profile.

DOI： 10.1088/1741-4326/aa4e58

記述言語：日本語   掲載種別：研究論文（その他学術会議資料等）  

記述言語：英語  

Experimental Advanced Superconducting Tokamak (EAST) aims to demonstrate steady-state advanced high-performance H-mode plasmas with an ITER-like configuration, plasma control and heating schemes. The plasma-facing components in EAST are actively cooled, providing good conditions for researching long-pulse and high-energy discharges. A long-pulse high-performance plasma discharge (#59892 discharge) of up to 103 s with a core electron temperature of up to 4.5 keV was sustained with an injected energy exceeding 0.22 GJ in the 2015-2016 experimental campaign. A calorimetric measurement utilizing the temperature increment of cooling water is carried out to calculate the heat load on the strike point region of the lower divertor during long-pulse discharges in EAST. For the long-pulse and high-energy discharges, the comparison of the measurement results for the heat load measured by divertor Langmuir probes and the calorimetry diagnostic indicates that most of the heat load is delivered to the divertor panels as plasma, not radiation, and charge exchange neutrals. The ratio of the heat load on the strike point region of the lower divertor to the total injected energy is on average 42.5&#37; per discharge with the lower single null divertor configuration. If the radiated energy loss measured by the fast bolometer diagnostic is taken into consideration, the ratio is found to be 61.6&#37;. The experimental results and the analysis of the physics involved in these discharges are reported and discussed.

DOI： 10.1088/1361-6587/aa5d88

記述言語：英語  

In the last few years, long-pulse H-mode plasma discharges (with small edge-localized modes and normalized beta, β _N ∼ 1) have been realized at the Experimental Advanced Superconducting Tokamak (EAST). This paper reports on high-β _N (>1.5) discharges in the 2015 EAST campaign. The characteristics of these H-mode plasmas have been presented in a database. Analysis of the experimental limit of β _N has revealed several main features of typical discharges. Firstly, efficient, stable high heating power is required. Secondly, control of impurity radiation (partly due to interaction between the plasma and the in-vessel components) is also a critical issue for the maintenance of high-β _N discharges. In addition an internal transport barrier (ITB) has recently been observed in EAST, introducing further improvement in confinement surpassing H-mode plasmas. ITB dynamics is another key issue for high-β _N plasmas in EAST. Each of these features is discussed in this paper. Study and improvement of these issues could be considered as the key to achieving long-pulse high-β _N operation with EAST.

DOI： 10.1088/1741-4326/aa626c

記述言語：英語  

Coaxial Helicity Injection (CHI) has now been implemented in QUEST. The goals for the first transient CHI experiments were to establish reliable gas breakdown conditions, and to measure CHI-produced toroidal current generation. Both these objectives were successfully met. Toroidal currents up to 29 kA were measured. Interestingly, these first plasmas on QUEST also suggest the formation of small amounts of closed magnetic flux surfaces.

DOI： 10.1585/pfr.12.1202020

記述言語：英語  

A typographical error of spelling occurred in author list of the published article. The correct author list is described as above on this page.

DOI： 10.1585/pfr.12.1902026

記述言語：英語  

Hydrogen plasma-driven permeation (PDP) experiments have been performed using a sputter-deposited tungsten (SP-W) coated F82H membrane in the spherical tokamak QUEST. It has been found that SP-W coatings tend to enhance hydrogen PDP compared with that of bare F82H membrane. Surface recombination is a key process determining the PDP flux, suggesting that surface effects on hydrogen PDP should be further investigated.

DOI： 10.1585/pfr.12.1305038

記述言語：英語  

DOI： 10.1016/j.fusengdes.2015.11.007

記述言語：英語  

DOI： 10.1016/j.fusengdes.2016.04.016

記述言語：英語  

DOI： 10.1016/j.fusengdes.2015.11.007

記述言語：日本語  

DOI： 10.1016/j.fusengdes.2015.12.022

記述言語：英語  

記述言語：英語  

The plasma control system (PCS) of QUEST is developed according to the progress of QUEST project. Since one of the critical goals of the project is to achieve the steady-state operation with high temperature vacuum vessel wall, the PCS is also required to have the capability to control the plasma for a long period. For the increase of the loads to processing power of the PCS, the PCS is decentralized with the use of reflective memories (RFMs). The PCS controls the plasma edge position with the real-time identification of plasma current and its position. This identification is done with not only flux loops but also hall sensors. The gas fueling method by piezo valve with monitoring the Hα signal filtered by a digital low-pass filter are proposed and suitable for the steady-state operation on QUEST. The present status and prospect of the PCS are presented with recent topics.

DOI： 10.1016/j.fusengdes.2016.04.016

記述言語：英語  

DOI： 10.1016/j.fusengdes.2015.11.035

記述言語：英語  

DOI： 10.1016/j.fusengdes.2015.12.022

記述言語：英語  

The high power and long pulse operation of the gyrotron as well as efficient transmission of its output are quite important for achieving improved plasma performances. A 28 GHz 1 MW gyrotron developed for GAMMA 10/PDX achieved an output power of 1.38 MW in 2015 experiment after the power supply was improved. Furthermore, a new 28/35 GHz dual-frequency gyrotron (2 MW 3 s and 0.4 MW CW) for QUEST, NSTX-U, Heliotron J and GAMMA 10/PDX has been fabricated, after the preliminary test of a double-disk sapphire window installed in the gyrotron was performed. In the first experimental test, the oscillation of the main mode was confirmed at a frequency of 28.036 GHz with a Gaussian-like beam and an output power of 1.22 MW.

DOI： 10.1063/1.4964176

記述言語：英語  

In order to evaluate the magnetic surface current density distributions, the electron guiding center orbits were calculated with the equilibrium fitting plasma equilibrium analysis. The current density distributions in electron cyclotron-heated plasma were estimated from the orbit analyses in the equilibrium plasma shaping for the fundamental and the second-harmonic resonant electrons separately. The current density distribution profiles on the equatorial plane were obtained for the electrons with initially positive and negative velocities parallel to the magnetic field direction as well as the trapped electrons. The surface averaged current density profiles of the closed flux surfaces were also evaluated. A significant amount of the positive current density distributions appeared outside the last closed flux surface (LCFS), while all the negative current density distributions were evaluated inside the LCFS. The trapped electrons being the second-harmonic electron cyclotron resonant contributed the negative current inside the LCFS. The positive current density distributions outside the LCFS were significant for the driven current density for both the fundamental and second-harmonic resonances.

DOI： 10.1109/TPS.2016.2570815

記述言語：英語  

The response of the far scrape-off layer (far-SOL) to strong gas puffing (SGP), and its role as the boundary condition for core plasma, are investigated using a two-point Langmuir probe measurement in the high poloidal beta configuration in the QUEST spherical tokamak. The temperature and heat flux behave in an opposite way in the far-SOL and end-plate region after SGP, although SGP increases the density globally. The apparent density decay time in the far-SOL area is much longer than that in the core. Significant co-current flow is driven solely by the electron cyclotron wave in the far-SOL flow. Sheared flow is also observed in the perpendicular velocity profile during the recovered current flat-top phase, and such flow profiles are flattened by SGP. These flow profiles are attributed not only to drift-driven flow but also to transport-driven flow, the sink effect on the end-plate, and the balance of the neutral particle source.

DOI： 10.1088/0741-3335/58/11/115004

記述言語：英語  

Fluctuations are measured in the edge and scrape-off layer (SOL) of QUEST using fast visible imaging diagnostic. Electron cyclotron wave injection in the Ohmic plasma features excitation of low frequency coherent fluctuations near the separatrix and enhanced cross-field transport. Plasma shifts from initial high field side limiter bound (inboard limited, IL) towards inboard poloidal null (IPN) configuration with steepening of the density profile at the edge. This may have facilitated the increased edge and SOL fluctuation activities. Observation of the coherent mode, associated plasma flow, and particle out-flux, for the first time in the IPN plasma configuration in a spherical tokamak may provide further impetus to the edge and SOL turbulence studies in tokamaks.

DOI： 10.1063/1.4960117

記述言語：英語  

Poloidal and radial distributions of the hydrogen flux, both atomic and ion, to the plasma facing materials (PFMs) are measured with four fixed and one movable PdCu membrane probes in the QUEST spherical tokamak. Direct information about hydrogen retention in the areas far from main plasma-wall interaction is essential because the significant part of the global retention is due to atomic hydrogen. Incident hydrogen flux Γ_inc is numerically reconstructed by fitting the experimentally measured hydrogen flux Γ_p, permeated through the membrane. The sensitivity of the probe is improved when compared with other work which allows dynamic changes of the retention flux to be monitored, allowing detection of 10&#37; of the Γ_inc modulation, or ∼10¹⁷ H m⁻²s⁻¹.

DOI： 10.1016/j.vacuum.2016.04.025

記述言語：英語   掲載種別：研究論文（国際会議プロシーディングス）  

記述言語：英語  

We investigate and calculate particle orbits and the effect of particle orbits on plasma current density for nonrelativistic resonance condition in the present RF-driven divertor plasma on QUEST. We surveyed particle orbits for different values of parallel refractive index, particle initial positions and pitch angles on fundamental and second harmonic resonance conditions. We observed that for fundamental harmonic resonance condition when particle orbits are plotted on the poloidal cross-section for positive values of parallel refractive index, these orbits are started from the resonance surface and produced their orbits around the LCFS (Large Closed Flux Surface). These orbits carry positive current. When particle orbits are plotted for negative values of parallel refractive index, these orbits are started from resonance surface, but remained at the inside of the LCFS. These orbits carry negative current that reduced the overall plasma current. For second harmonic resonance condition when particle orbits are plotted on the poloidal cross-section most of the orbits remained in inside the LCFS and carry positive current. When we consider the value of parallel refractive index-0.4 and +0.4 some particle orbits arrived at the limiter and become lost particles. On the other hand, when we consider particle initial positions 0.16 m or more vertically far from the mid plane some banana orbits are produced. These banana orbits make the current density profile maximum at low field side region. From this calculation we got a hollow current density profile with current density peak at the low field side region outside of the LCFS. From this calculation we can infer that parabolic current density profile is possible, if we set the resonance surface outside of the magnetic axis by increasing the toroidal magnetic field coil current and make the plasma position inward by increasing vertical field coil current.

DOI： 10.1109/SOFE.2015.7482308

記述言語：英語  

Measurement of the Electron Cyclotron Emission (ECE) spectrum is one of the most popular electron temperature diagnostics in nuclear fusion plasma research. A 2-dimensional ECE imaging systemwas developed with an adaptive-Array approach. Aradio-frequency (RF) heterodyne detection system with Software Defined Radio (SDR) devices and a phased-Array receiver antenna was used to measure the phase and amplitude of the ECE wave. The SDR heterodyne system could continuously measure the phase and amplitude with sufficient accuracy and time resolution while the previous digitizer system could only acquire data at specific times. Robust streaming phase measurements for adaptive-Arrayed continuous ECE diagnostics were demonstrated using Fast Fourier Transform (FFT) analysis with the SDR system. The emission field pattern was reconstructed using adaptive-Array analysis. The reconstructed profileswere discussed using profiles calculated from coherent single-frequency radiation from the phase array antenna.

DOI： 10.1088/1748-0221/11/04/C04010

記述言語：英語  

Spontaneous toroidal rotation of the plasma is observed in the spherical tokamak QUEST with the help of electron cyclotron resonance (ECR) heating and without the use of any externally injected momentum. Several vertical magnetic field ( B_z) configurations with varying mirror ratio (M) (a measure of field curvature) are applied and evolution of rotation is studied with the help of Doppler spectroscopy of bulk and impurity ions. Significant toroidal rotation (V_φ-6km/s) is initiated in the open magnetic field configuration during the initial plasma breakdown phase, which is later sustained ( V_φ- 20km/s) in a closed magnetic field configuration in steady state. Rotational velocity is primarily along the cocurrent direction and is found to be proportional to the B_z strength and the resulting plasma current. High M and B_z are demonstrated to be the two specific external controls by which rotation can be initiated in the plasma. The rotation in open field lines is found to be initiated at the ECR layer in the slablike plasma, which is evolved to produce a sustained rotation in the natural divertor inboard poloidal field null equilibrium in QUEST.

DOI： 10.1109/TPS.2016.2522765

記述言語：英語  

DOI： 10.1088/1748-0221/11/01/C01014

記述言語：英語  

DOI： 10.1088/1748-0221/11/01/C01049

記述言語：英語  

DOI： 10.1016/j.fusengdes.2015.06.054

記述言語：英語  

DOI： 10.1016/j.jnucmat.2014.11.136

記述言語：英語  

記述言語：英語  

Heat flux and plasma flow in the scrape-off layer (SOL) are examined for the inboard poloidal field null (IPN) configuration of the spherical tokamak QUEST. In the plasma current (I_p) ramp-up phase, high heat flux (>1 MW/m²) and supersonic flow (Mach number M > 1) are found to be present simultaneously in the far-SOL. The heat flux is generated by energetic electrons excursed from the last closed flux surface. Supersonic flows in the poloidal and toroidal directions are correlated with each other. In the quasi-steady state, sawtooth-like oscillation of I_p at 20 Hz is observed. Heat flux and subsonic plasma flow in the far-SOL are modified corresponding to the I_p-oscillation. The heat flow caused by motion of energetic electrons and the bulk-particle transport to the far-SOL is enhanced during the low-I_p phase. Modification of plasma flow in the far SOL occurs earlier than the I_p crash. The M-I_p curve has a limit-cycle characteristic with sawtooth-like oscillation. Such a core-SOL relationship indicates that the far-SOL flow plays an important role in sustaining the oscillation of I_p in the IPN configuration.

DOI： 10.1063/1.4928878

記述言語：英語  

DOI： 10.1016/j.jnucmat.2014.12.092

記述言語：英語  

記述言語：英語  

記述言語：英語  

Abstract Global particle balance in non-inductive long-duration plasma on QUEST has been investigated. Approximately 70&#37; of the fuel hydrogen (H) was retained in the wall and then was almost exhausted just after the discharge. The global recycling ratio (R_g), defined as the ratio of the evacuated H₂ flux to that injected, was found to gradually increase during discharges and subsequently rose rapidly. To study the growth of R_g, the thermal desorption spectra after deuterium implantation in a specimen exposed to QUEST plasma was analyzed with a model which includes reflection, diffusion, solution, recombination, trapping, and plasma-induced desorption in the re-deposition layer. The model reconstructs the growth of R_g during a long-duration plasma and indicates solution plays a dominant role in the growth.

DOI： 10.1016/j.jnucmat.2015.01.013

記述言語：英語  

At the Plasma Research Center at University of Tsukuba, development of megawatt gyrotrons is being performed as a collaborative electron cyclotron heating (ECH) study with some research organizations. A 28 GHz 1 MW 1 s gyrotron has been developed to upgrade the GAMMA 10/PDX ECH systems. To improve the oscillation efficiency in high current regions, the magnetron injection gun (MIG) of the 28 GHz gyrotron has been modified. Output power of 1.25 MW has been achieved with this gyrotron. For the first step of the collaborative research between Tsukuba University and Kyushu University, the Tsukuba 28 GHz gyrotron was adapted to the Q-shu University Experiment with Steady-State Spherical Tokamak (QUEST) ECH system, and the plasma heating and current drive effect were demonstrated. We obtained successful results, including an electron cyclotron-driven plasma current of 66 kA in the QUEST plasma experiment. For the next step of the collaborative research, the design targets of a 2 MW 3 s and 0.4 MW continuous wave have been achieved in a design study of a new 28 GHz gyrotron.

DOI： 10.13182/FST14-848

記述言語：英語  

記述言語：英語  

An orthogonal dynamic programming (ODP) based particle image velocimetry (PIV) technique is developed to measure the time resolved flow field of the fluctuating structures at the plasma edge and scrape off layer (SOL) of tokamaks. This non-intrusive technique can provide two dimensional velocity fields at high spatial and temporal resolution from a fast framing image sequence and hence can provide better insights into plasma flow as compared to conventional probe measurements. Applicability of the technique is tested with simulated image pairs. Finally, it is applied to tangential fast visible images of QUEST plasma to estimate the SOL flow in inboard poloidal null-natural divertor configuration. This technique is also applied to investigate the intricate features of the core of the run-away dominated phase following the injection of a large amount of neutrals in the target Ohmic plasma. Development of the ODP-PIV code and its applicability on actual plasma images is reported.

DOI： 10.1063/1.4914838

記述言語：英語  

An orthogonal dynamic programming (ODP) based particle image velocimetry (PIV) technique is developed to measure the time resolved flow field of the fluctuating structures at the plasma edge and scrape off layer (SOL) of tokamaks. This non-intrusive technique can provide two dimensional velocity fields at high spatial and temporal resolution from a fast framing image sequence and hence can provide better insights into plasma flow as compared to conventional probe measurements. Applicability of the technique is tested with simulated image pairs. Finally, it is applied to tangential fast visible images of QUEST plasma to estimate the SOL flow in inboard poloidal null–natural divertor configuration. This technique is also applied to investigate the intricate features of the core of the run-away dominated phase following the injection of a large amount of neutrals in the target Ohmic plasma. Development of the ODP-PIV code and its applicability on actual plasma images is reported.

記述言語：英語  

The method of plasma current profile reconstruction using the polarimeter/interferometer (POINT) data from a simulated equilibrium is explored and validated. It is shown that the safety factor (q) profile can be generally reconstructed from the external magnetic and POINT data. The reconstructed q profile is found to reasonably agree with the initial equilibriums. Comparisons of reconstructed q and density profiles using the magnetic data and the POINT data with 3&#37;, 5&#37; and 10&#37; random errors are investigated. The result shows that the POINT data could be used to a reasonably accurate determination of the q profile.

DOI： 10.1088/1009-0630/17/1/14

記述言語：英語  

記述言語：英語  

High poloidal beta (εβ_p ~ 1) operation in steady state condition in tokamaks is of great interest and has previously been demonstrated using NBI, LHCD and low current (I_p) plasma for a short time (<0.5 s). A very few experiments however, have been performed towards the investigation of highest obtainable β_p in tokamak plasma. In this work we report the first result of high β_p production and its sustainment though an off axis ECCD at two different frequencies (fundamental and second harmonic) in Ohmic (OH) target plasma. With application of ECCD, plasma β_p increased to encounter an equilibrium limit and the standard limiter configuration is transformed to an Inboard Poloidal field Null (IPN) configuration. Both off-axis and on-axis ECCD is studied and found to have some distinctive features, which are discussed in this paper.

DOI： 10.1051/epjconf/20158702012

記述言語：英語  

Comparison between non-inductive plasma current start-up using electron cyclotron resonance heating (ECRH) with and without fundamental (1st) resonance using 8.2-GHz microwave is experimentally performed on QUEST. Results of soft X-ray measurements in different energy ranges indicate significant difference in the number of current-carrying electrons with and without fundamental resonance. Numerical calculations show effective extendibility of electron energy in ECRH of fundamental resonance.

DOI： 10.1585/pfr.10.3402066

記述言語：英語  

記述言語：英語  

DOI： 10.1016/j.jnucmat.2014.12.092

記述言語：英語  

A thermal imaging system to measure plasma Electron Bernstein Emission (EBE) emanating from the mode conversion region in overdense plasma is discussed. Unlike conventional ECE/EBE imaging, this diagnostics does not employ any active mechanical scanning mirrors or focusing optics to scan for the emission cones in plasma. Instead, a standard 3 * 3 waveguide array antenna is used as a passive receiver to collect emission from plasma and imaging reconstruction is done by accurate measurements of phase and intensity of these signals by heterodyne detection technique. A broadband noise source simulating the EBE, is installed near the expected mode conversion region and its position is successfully reconstructed using phase array technique which is done in post processing.

DOI： 10.1063/1.4889903

記述言語：英語  

A Phased Array Antenna (PAA) was considered as launching and receiving antennae in reflectometry to attain good directivity in its applied microwave range. A well-focused beam was obtained in a launching antenna application, and differential-phase evolution was properly measured by using a metal reflector plate in the proof-of-principle experiment at low power test facilities. Differential-phase evolution was also evaluated by using the PAA in the Q-shu University Experiment with Steady State Spherical Tokamak (QUEST). A beam-forming technique was applied in receiving phased-array antenna measurements. In the QUEST device that should be considered as a large oversized cavity, standing wave effect was significantly observed with perturbed phase evolution. A new approach using derivative of measured field on propagating wavenumber was proposed to eliminate the standing wave effect.

DOI： 10.1063/1.4901018

記述言語：英語  

A Phased Array Antenna (PAA) was considered as launching and receiving antennae in reflectometry to attain good directivity in its applied microwave range. A well-focused beam was obtained in a launching antenna application, and differential-phase evolution was properly measured by using a metal reflector plate in the proof-of-principle experiment at low power test facilities. Differential-phase evolution was also evaluated by using the PAA in the Q-shu University Experiment with Steady State Spherical Tokamak (QUEST). A beam-forming technique was applied in receiving phased-array antenna measurements. In the QUEST device that should be considered as a large oversized cavity, standing wave effect was significantly observed with perturbed phase evolution. A new approach using derivative of measured field on propagating wavenumber was proposed to eliminate the standing wave effect.

DOI： 10.1063/1.4891710

記述言語：英語  

記述言語：英語  

Formations of a closed flux surface (CFS) on QUEST are achieved by fully non-inductive current start-up driven by RF, which is 8.2GHz in frequency and more than 40kW in power. It found that appropriate magnetic configuration with positive n-index and reduction of particle recycling was crucial to achieve the non-inductive plasma current start-up (PCS) successfully. Especially the controllability of particle recycling should be improved by wall conditioning based on successive plasma production and wall cleaning with electron cyclotron resonance heating (ECR) plasmas induced by RF in frequency of 2.45GHz.

DOI： 10.1088/1742-6596/511/1/012041

記述言語：英語  

In the present RF-driven plasma with a lot of high-energy electrons, there may be anisotropic plasma pressure, which makes the usual equilibrium analysis difficult, but the Cauchy condition surface method can reconstruct the plasma shape precisely regardless of the anisotropy. In addition, the plasma current effect in the open magnetic surfaces outside of the closed magnetic surfaces is considered in the RF-driven divertor plasma. In the reconstruction process, singular value (SV) decomposition is used and optimal criterion function for generalized cross validation is estimated concerning truncation or reduction of the small-SV components.

DOI： 10.1109/TPS.2014.2341616

記述言語：英語  

Statistical features of fluctuations are investigated using the fast camera imaging technique in the scrape of layer (SOL) of electron cyclotron resonance heated Ohmic plasma. Fluctuations in the SOL towards low field side are dominated by coherent convective structures (blobs). Two dimensional structures of the higher order moments (skewness s and kurtosis k) representing the shape of probability density function (PDF) are studied. s and k are seen to be functions of the magnetic field lines. s and k are consistently higher towards the bottom half of the vessel in the SOL showing the blob trajectory along the field lines from the top towards bottom of the vessel. Parabolic relation (k = As-2 + C) is observed between s and k near the plasma boundary, featuring steep density gradient region and at the far SOL. The coefficient A, obtained experimentally, indicates a shift of prominence from pure drift-wave instabilities towards fully developed turbulence. Numerical coefficients characterizing the Pearson system are derived which demonstrates the progressive deviation of the PDF from Gaussian towards gamma from the density gradient region, towards the far SOL. Based on a simple stochastic differential equation, a direct correspondence between the multiplicative noise amplitude, increased intermittency, and hence change in PDF is discussed. (C) 2014 AIP Publishing LLC.

DOI： 10.1063/1.4890359

記述言語：英語  

High βp (εβp ∼ 1) equilibria obtained in a ECW heated Ohmic plasma is investigated using a simple analytic solution of Grad-Shafranov equation. The formation of a natural inboard poloidal null associated with high βp is explained consistently by high diamagnetism and negative triangularity. As βp is increased, the poloidal null point penetrates further into the vacuum vessel, which is qualitatively explained by the analytic model. Transition from inboard (high field side) limiter bound to the natural divertor configuration is associated with a reduction of the edge safety factor without appreciable enhancement of MHD activities. Such a scenario is also addressed successfully with the model.

DOI： 10.1585/pfr.9.3402093

記述言語：英語  

記述言語：英語  

A scenario for non-inductive current ramp-up has been demonstrated using electron cyclotron waves in the spherical tokamak QUEST. The configuration was characterized by a high toroidal magnetic mirror ratio of 2 and a steady vertical magnetic field of more than 10&#37; of the toroidal magnetic field. The generation and confinement of energetic electrons having energy greater than 10 keV were studied using hard x-rays. Because of the energetic electron pressure, a natural divertor formed with an inboard poloidal field null at the high poloidal beta (approximately 3-4).

DOI： 10.1088/0029-5515/54/2/023010

記述言語：英語  

A drift error correction technique with machine vision and a real-time equilibrium calculation code have been developed on the QUEST (Q-shu university experiment with the steady-state spherical tokamak) for steady-state operation. The drift error caused by the long time-integration of magnetic raw signals has to be removed. With a captured image of the plasma’s cross section, the plasma’s position is identified by use of image filters. The measured magnetic flux values are corrected to the calculated flux values estimated by using this plasma position. The correction with the captured image work as expected in the preliminary result using a flashlight instead of a plasma.

DOI： 10.3938/jkps.65.1191

記述言語：英語  

A solution of two-fluid (electron and ion), axisymmetric equilibrium is presented that approximates solenoidfree plasmas sustained only by RF electron heating that are recently studied in TST-2, LATE, QUEST. These plasmas indicate presence of orbit-confined energetic electrons carrying substantial toroidal current outside the last closed flux surface (LCFS); T_e/Ti » 1 and low collisionality at modest densities within LCFS; and likely a positive plasma potential relative to the conductive vacuum vessel. A system of nonlinear second-order partial differential and algebraic equations constraining six functionals of poloidal magnetic flux or canonical angular momentum are solved. An example plasma measured in TST-2 is used to guide, by trial and error, the selection of these functionals to find appropriate solutions, while assuming peaked plasma profiles and 60&#37; toroidal current within the LCFS. The numerical equilibrium obtained indicates a substantial ion toroidal flow and electrostatic potential so that the ion ∇p_i, centrifugal, and electrostatic forces of nearly equal magnitudes combine to balance the J_i × B force, differently from the massless electron fluid that satisfies ∇p_e = J_e × B. The calculated properties suggest additional measurements needed to refine the choices of the functional forms and improve the two-fluid equilibrium fit to such plasmas.

DOI： 10.1585/pfr.9.3403146

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

In the present RF-driven plasma with a lot of high-energy electrons, there may be anisotropic plasma pressure, which makes difficult a usual equilibrium analysis, but the CCS method can reconstruct the plasma shape precisely regardless of the anisotropy. And the plasma current effect in the open magnetic surfaces outside of the closed magnetic surfaces is considered in the RF-driven plasma. Further high-energy particle guiding center orbits are calculated aiming at estimation of the open-magnetic-surface current distribution.

DOI： 10.1109/SOFE.2013.6635507

記述言語：英語  

A plasma control system to sustain divertor configurations is developed on QUEST (Q-shu university experiment with steady-state spherical tokamak). Magnetic fluxes are numerically integrated at 100 kHz using FPGA (Field-Programmable Gate Array) modules and transferred to a main calculation loop at 4 kHz. With these signals, plasma shapes are identified in real time at 2 kHz under the assumption that the plasma current can be represented as one filament current. This calculation is done in another calculation loop in parallel by taking advantage of a multi-core processor of the plasma control system. The inside and outside plasma edge positions are controlled to their target positions using PID (proportional-integral-derivative) control loops. Whereas the outside edge position can not be controlled by the outer PF coil current, the inside edge position can be controlled by the inner PF coil current. (C) 2013 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2013.03.035

記述言語：日本語  

記述言語：英語  

記述言語：英語  

To investigate the effects of confinement of fast electrons on poloidal beta equilibrium, experiments under different toroidal magnetic mirror ratios, MOMFC, were performed in electron cyclotron heated plasma in the spherical tokamak QUEST. The proportional constants of the relation between plasma current and vertical magnetic fields decreased with increase in MOMFC. The poloidal field null point appeared in the chamber for MOMFC > 1.8 at a plasma current value of ∼15 kA. Poloidal beta and temperature of hard X-rays increased with MOMFC. This suggests that high poloidal beta equilibrium plasma is produced by better confinement of energetic trapped particles.

記述言語：日本語  

記述言語：英語  

A particle image velocimetry technique based on orthogonal dynamic programming is developed to measure the time resolved flow field of the fluctuating structures at the plasma edge and scrape off layer. This non-intrusive technique can provide two dimensional velocity fields at high spatial and temporal resolution from a fast framing image sequence and hence can provide better insights in plasma flow as compared to conventional probe measurements. Applicability of the technique is tested with simulated image pairs. Finally, it is applied to tangential fast visible images of QUEST as a test case to estimate the scrape off layer flow in Ohmic and ECRH driven plasma discharges.

記述言語：英語  

Blob-filaments have been observed by combined measurement with a fast camera and a movable Langmuir probe in an open magnetic field line configuration of electron cyclotron resonance (ECR) heating plasma in QUEST. Blob-filaments extended along field lines do correspond to over-dense plasma structures and propagated across the field lines to the outer wall. The radial velocity of the blob structure, Vb, was obtained by three methods and was dominantly driven by the E x B force. The radial velocity, size of the blob showed good agreements with the results obtained by sheath-connected interchange theoretical model. Vb corresponds to roughly 0.02-0.07 of the local sound speed (Cs) in QUEST. The higher moments (skewness S and kurtosis K) representing the shape of PDF of density fluctuation are studied. Their least squares fitting with quadratic polynomial is K = (1.60 +/- 0.27) S-2 - (0.46 +/- 0.20). The larger blob structures, occurring only 10&#37; of the time, can carry more than 60&#37; loss of the entire radial particle flux. (C) 2013 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2013.01.106

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

QUEST has a divertor configuration with a high and a negative n-index, and the problem of plasma vertical position instability control in QUEST is still under extensive study for achieving high efficiency plasma. The instability we considered is that the toroidal plasma moves either up or down in the vacuum chamber until it meets the vessel wall and is extinguished. The actively controlled coils (HCU and HCL) outside the vacuum vessel are serially connected in feedback with a measurement of the plasma vertical position to provide stabilizing control. In this work, a robust controller is employed by using the loop synthesis method, and provides robust stability over a wide range of n-index. Moreover, the gain of the robust controller is lower than that of a typical proportional derivative (PD) controller in the operational frequency range; it indicates that the robust controller needs less power consumption than the PD controller does.

DOI： 10.1088/1009-0630/15/3/21

記述言語：英語  

To investigate the effects of confinement of fast electrons on poloidal beta equilibrium, experiments under different toroidal magnetic mirror ratios, MOMFC, were performed in electron cyclotron heated plasma in the spherical tokamak QUEST. The proportional constants of the relation between plasma current and vertical magnetic fields decreased with increase in MOMFC. The poloidal field null point appeared in the chamber for MOMFC > 1.8 at a plasma current value of ~15 kA. Poloidal beta and temperature of hard X-rays increased with MOMFC. This suggests that high poloidal beta equilibrium plasma is produced by better confinement of energetic trapped particles.

DOI： 10.1585/pfr.8.2402118

記述言語：英語  

A particle image velocimetry technique based on orthogonal dynamic programming is developed to measure the time resolved flow field of the fluctuating structures at the plasma edge and scrape off layer. This nonintrusive technique can provide two dimensional velocity fields at high spatial and temporal resolution from a fast framing image sequence and hence can provide better insights in plasma flow as compared to conventional probe measurements. Applicability of the technique is tested with simulated image pairs. Finally, it is applied to tangential fast visible images of QUEST as a test case to estimate the scrape off layer flow in Ohmic and ECRH driven plasma discharges.

DOI： 10.1585/pfr.8.2402098

記述言語：英語  

Fluctuations in slab plasma produced by electron cyclotron waves in an open magnetic configuration with vertical (B-z) and toroidal (B-t) fields are investigated as a function of B-z/B-t (field pitch) using a high speed visible camera in the spherical tokamak QUEST. Higher order moments (skewness s and kurtosis k) of spatio- temporal fluctuations are investigated. Background fluctuations, generation of coherent convective structures (blobs) and their propagation are analysed at the intensity gradient (R-im) and the source- free (essentially vacuum on the low- field side) regions, respectively. Parabolic relation (k = As-2 + C) is observed between s and k on the entire R- Z plane for a wide range of B-z/B-t. Progressive shift from the Gaussian statistics beyond R-im is observed with increasing B-z/B-t. From the moments of the probability density functions, possible location for blob generation is inferred to be R-im. Direct correspondence is observed with the prominence of the blobs and increase in stochastic forces with the increase in field pitch. Distribution of waiting time suggests that the blob generation may be a Poisson process. Accelerated cross field transport is observed for reasonably large-sized blobs.

DOI： 10.1088/0029-5515/52/12/123016

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

記述言語：英語  

A CW phased-array antenna system for electron cyclotron/Bernstein wave heating and current drive (ECWH/CD, EBWH/CD) experiments was developed in the QUEST. The antenna was designed to excite an elliptically polarized pure O-mode wave in oblique injection for the O-X-B mode conversion scenario, and its good performance was confirmed at a high power level. Long pulse discharges with a plasma current of 10 kA and 15 kA were non-inductively attained for 37 s and 20 s, respectively, with only radio frequency (RF) power. Divertor configurationswere also obtained in the RF-sustained plasmas. A new operational window for sustained plasma current was observed in the high-density plasma with a higher RF incident power. Two new heating and current drive systems with an 8.56GHz klystron and a 28GHz gyrotron are being prepared to conduct CW EBWH/CD experiments in the high-density plasma.

DOI： 10.1585/pfr.7.2402112

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

Ambient fluctuations and the generation of coherent convective structures are investigated as a function of vertical to toroidal magnetic field ratio (B _z/B _t) in the spherical tokamak QUEST [K. Hanada et al., Plasma Fusion Research 5, S1007]. Fluctuation signals are obtained at a high spatial and temporal resolution using the fast visible camera. The coherent structures (blobs) were seen to be apparent only beyond a critical Bz/Bt ratio (∼ 10 ^-2). This clearly indicates a shift from the drift to the interchange mode with the increase in vertical magnetic field. The blobs are propagated radially at a high speed ∼ 10 ³ m/s which is ∼ 1/10 ^th of the ion acoustic speed in QUEST. Lifetime of the blobs is about 700 μs. Second part of the study comprises of the investigation of the finer structures in the current ramp-up phase which is characterized by lower levels of fluctuations. Singular value decomposition (SVD) and principal component analysis (PCA) techniques are applied on the fast camera images to reveal the structures which are otherwise embedded in the overall intensity image. Electron banana orbits were seen to form from the slab annular plasma and grow wider to eventually shape the plasma current front.

DOI： 10.1541/ieejfms.132.545

記述言語：英語  

Recent progress towards SSO of a spherical tokamak (ST) on QUEST is introduced. Zero-dimensional calculations of particle balance for QUEST experiment, refresh wall, and hot wall were performed showing that a SSO time of more than 1000s is required to obtain reliable results in regard to particle balance. Basic features of blobs were investigated in electron cyclotron resonance (ECR) heated plasmas. Formation and maintenance of a single-null divertor configuration was successfully obtained. Plasma duration in the divertor configuration was limited by out-gassing from a non-water-cooled first wall. Design studies for a hot wall are introduced.

DOI： 10.1541/ieejfms.132.490

記述言語：英語  

The power deposition profiles were analyzed with a multiple ray tracing code for the Electron Bernstein Wave Heating and Current Drive (EBWH/CD) experiments in QUEST. In the EBWH/CD experiments in QUEST, the OXB mode conversion scenario was selected for the plasma current sustainment in the rather low-density case. The algorithm for the wave penetration through evanescent layer beyond a O-mode cutoff position was developed for the multiple-ray analysis. The launching antenna positions were considered to obtain the significant wave absorption in the specific propagating direction for current drive experiments using the developed ray-tracing code.

DOI： 10.1541/ieejfms.132.505

記述言語：英語  

Electron Bernstein Wave Heating and Current Drive (EBWH/CD) experiments have been conducted to sustain a spherical tokamak configuration in a steady state in QUEST. In the EBWH/CD experiments on QUEST, the O-X-B mode conversion scenario was selected for plasma current sustainment in rather lowdensity plasmas. A new phased-array antenna system was proposed to launch a pure elliptically-polarized O-mode in the oblique injection. The prototype antenna system was designed and its performance was checked at low power test facilities. Good focusing and steering properties were confirmed in the low power test.

DOI： 10.1541/ieejfms.132.511

記述言語：英語  

In order to identify plasma shape, there is a way to represent the plasma current profile with several parameters, and adjust these parameters with least-square technique in order for calculated magnetic values to accord with measured ones. Here, the plasma shape parameters such as minor radius, elongation, and triangularity are chosen as the fitting parameters to represent plasma shape more directly, and the applicability to the control of the plasma shape are described by evaluating its calculation time. In order to find minimum of an objective function with least-square technique, two methods are compared, namely a linear approximation method and a downhill simplex method. While high accuracies of the measured magnetic signals are required, the good reproducibility is obtained, and the plasma shape identification can be done within several milliseconds in both methods.

DOI： 10.1541/ieejfms.132.477

記述言語：英語  

A fast visible imaging system is installed on the spherical tokamak QUEST to study edge turbulence. The camera uses a complementary metal-oxide semiconductor detector with a maximum resolution of 1024 x 1024 at 7000 frames/s (fps) and can achieve 775 kfps at a resolution of 128 x 24. In this paper, we present the salient features of the system and its application to study edge turbulence in 8.2 GHz ECRH driven slab plasma, without plasma current. Vertical magnetic field (B-z) topology is varied with three sets of poloidal field (PF) coils and the variation in the edge turbulence is investigated as a function of the B-z strength and curvature. Fluctuation amplitude was highest for the shallow PF well. Cross-correlation coefficient shows distinct coherent mode along z direction at the steep density gradient region and it grows with the PF mirror ratio. (C) 2012 American Institute of Physics. [http://dx.doi.org/10.1063/1.4739080]

DOI： 10.1063/1.4739080

記述言語：英語  

記述言語：英語  

記述言語：英語  

A CW phased-array antenna system for electron cyclotron/Bernstein wave heating and current drive (ECWH/CD, EBWH/CD) experiments was developed in the QUEST. The antenna was designed to excite an elliptically polarized pure O-mode wave in oblique injection for the O-X-B mode conversion scenario, and its good performance was confirmed at a high power level. Long pulse discharges with a plasma current of 10 kA and 15 kA were non-inductively attained for 37 s and 20 s, respectively, with only radio frequency (RF) power. Divertor configurations were also obtained in the RF-sustained plasmas. A new operational window for sustained plasma current was observed in the high-density plasma with a higher RF incident power. Two new heating and current drive systems with an 8.56 GHz klystron and a 28 GHz gyrotron are being prepared to conduct CW EBWH/CD experiments in the high-density plasma.

記述言語：英語  

記述言語：英語  

記述言語：日本語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

Electron Bernstein Wave Heating and Current Drive (EBWH/CD) experiments have been conducted to sustain a spherical tokamak configuration in a steady state in QUEST. In the EBWH/CD experiments on QUEST, the O-X-B mode conversion scenario was selected for plasma current sustainment in rather low-density plasmas. A new phased-array antenna system was proposed to launch a pure elliptically-polarized O-mode in the oblique injection. The prototype antenna system was designed and its performance was checked at low power test facilities. Good focusing and steering properties were confirmed in the low power test.

記述言語：英語  

In order to identify plasma shape, there is a way to represent the plasma current profile with several parameters, and adjust these parameters with least-square technique in order for calculated magnetic values to accord with measured ones. Here, the plasma shape parameters such as minor radius, elongation, and triangularity are chosen as the fitting parameters to represent plasma shape more directly, and the applicability to the control of the plasma shape are described by evaluating its calculation time. In order to find minimum of an objective function with least-square technique, two methods are compared, namely a linear approximation method and a downhill simplex method. While high accuracies of the measured magnetic signals are required, the good reproducibility is obtained, and the plasma shape identification can be done within several milliseconds in both methods.

記述言語：英語  

記述言語：英語  

記述言語：英語  

After several non-inductive current startup experimental campaigns in the spherical tokamak QUEST, its metallic first walls have revealed various kinds of damages as a signature of strong plasma wall interaction (PWI). Several types of footprint traces, namely colored regions formed due to material erosion/redeposition, melting of plasma facing components (PFCs) and numerous arc tracks on the chamber walls are recognized. Analysis of the re-deposited materials on collector probes is carried out using Xray photoelectron spectroscopy (XPS), scanning electron microscopy (SEM) and energy dispersive X-ray spectrometry (EDS). Redeposition of several impurity materials such as carbon, oxygen and tungsten is identified. The footprint traces are majorly formed on the lower side PFCs, showing a large up/down asymmetry. Both toroidally symmetric and asymmetric footprint traces are formed on the bottoms side divertor plate and the lower part of the outboard side walls, respectively. Localized melting occurred on the outboard side limiters is attributed to the loss of energetic electrons produced via electron cyclotron resonance (ECR) heating. The observed damages are discussed in view of localized PWI, loss of energetic electrons, particle drifts, sputtering, arcing and redeposition of eroded materials. Material analysis and numerically calculated guiding center orbits of the charge particles are used to discuss these damages. (C) 2011 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2011.10.001

記述言語：英語  

In order to investigate the overall atomic hydrogen background and the dynamic characteristics of wall pumping/fuelling phenomenon, a permeation probe system has been developed and applied in the spherical tokamak QUEST. Reliability of measurements, within +/- 3&#37; accuracy and a positive correlation with the hydrogen line emission over three orders of magnitude have been demonstrated for more than 3000 various plasma discharges. By comparison of the experimental permeation (flux) curves with the numerically simulated curves, the net incident atomic hydrogen flux is evaluated in the range of 1 x 10(19) H m(-2) s(-1) 4 x 10(20) H m(-2) s(-1). The atomic flux has been investigated as a function of various plasma operation parameters like RF power, gas pressure and magnetic configuration. Using the static particle balance and permeation measurements, the progress in wall conditioning has been investigated. An inverse correlation between the atomic hydrogen flux and improvement in wall pumping has been observed over the two campaigns. (C) 2011 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2011.09.026

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

The 16-element phased-array antenna has been developed for the EBWH/CD experiments. The 2-D k-spectra could be controlled. The heat load and stress were analyzed to be moderate in the 200kW CW operation. The symmetrically 2-D narrow-field distributions were excited by the developed antenna, indicating its good applicability to the EBWH/CD experiments.

DOI： 10.1109/irmmw-THz.2011.6105070

記述言語：英語  

記述言語：英語  

Because QUEST tokamak has a divertor configuration with a higher κ and a negative n-index, a precise power supply with a rapid response is needed to control the vertical position of the plasma. A matrix converter is a direct power conversion device that uses an array of controlled bidirectional switches as the main power elements for creating a variable-output current system. This paper presents a novel three-phase to two-phase topological matrix converter as a proposed power supply that stabilizes the plasma vertical position and achieves unity input power factor. An indirect control strategy in which the matrix converter is split into a virtual rectifier stage and a virtual inverter stage is adopted. In the virtual rectifier stage, the instantaneous active power and reactive power are decoupled on the basis of system equations derived from the DQ transformation; hence, unity power factor is achieved. Space vector pulse width modulation is adopted to determine the switching time of each switch in the virtual rectifier; the output voltage of the virtual rectifier is adjusted by the virtual inverter stage to obtain the desired load current. Theoretical analyses and simulation results are provided to verify its feasibility.

DOI： 10.1585/pfr.6.2405137

記述言語：英語  

Ohmic plasma currents of up to 17 kA with a discharge duration of 0.32 s have been obtained in the Kyushu University Experiment with Steady-State Spherical Tokamak (QUEST) with the help of electron cyclotron wave (ECW) and cancellation coils (CCs). The CCs, originally installed to create a field null in the plasma breakdown phase, are essential for producing plasma current in QUEST. Although the ohmic coil current is initially biased and then reduced completely to zero to induce the plasma current in 15-20 ms, we demonstrate that the flat top of the plasma current exceeding 20 ms is maintained by the vertical field after the ohmic current is switched off. This type of operation is quite favorable for extending pulsed operation to the steady state by electron Bernstein wave current drive (EBCD).

DOI： 10.1585/pfr.6.1402003

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

The phased-array antenna system for Electron Cyclotron/Bernstein Wave Heating and Current Drive experiments has been developed in the QUEST. The antenna was designed to excite a pure O-mode wave in the oblique injection for the O-X-B mode conversion experiments, and its good performances were confirmed at a low power level. The plasma current (<∼15kA) with an aspect ratio of 1.5 was started up and sustained by only RF injection in the low-density operations. The long pulse discharge of 10 kA was also attained for 37 s. The new density window to sustain the plasma current was observed in the high-density plasmas. The single-null divertor configuration with the high plasma current (<∼25kA) was attained in the 17 s plasma sustainment.

DOI： 10.1063/1.3665017

記述言語：英語  

A permeation measuring system with a nickel membrane of 30 ion thickness was installed near the mid plane of the spherical tokamak, QUEST Hydrogen permeation through the membrane heated at fix temperatures (422 - 506 K) was measured during short pulse (< 1 s) and long pulse (I hour) plasma discharges. After the membrane was heated to a required temperature, hydrogen plasma was discharged using a 2.45 GHz or 8.2 GHz RF system. Significant plasma-driven permeation was observed even for very short plasma discharges (e.g. 0.1s). Numerical calculations with the use of diffusion equation under recombination boundary conditions were conducted to simulate the transient permeation behavior. The numerical calculations were also used to estimate diffusion coefficient and recombination coefficients of membrane material. Temperature dependence of both the coefficients was explained by the Arrhenius law. A one hour long permeation curve was also numerically reproduced using the same set of parameters except an increasing recombination coefficient on the plasma side of the membrane.

記述言語：英語  

In magnetized plasmas, the presence of a significant number of energetic electrons has been observed but quantitative characteristics of these electrons are proving difficult to investigate. A Langmuir probe offers a means to provide quantitative measurement of these energetic electrons that takes into account electron emissions (secondary electron emission and electron reflection) from the probe tips and sheath expansion around the probe tips caused by a considerable negative potential. In this paper, these effects are experimentally confirmed and an analytical means to measure energetic electron characteristics are proposed. An analysis of plasmas produced by a high frequency wave is then applied leading to the successful detection of an asymmetric flow of energetic electrons. The estimated electron temperature and current density were approximately 4-5 keV and 2-3 kA/m(2). (C) 2011 American Institute of Physics. [doi:10.1063/1.3662204]

DOI： 10.1063/1.3662204

記述言語：英語  

記述言語：英語  

Edge turbulence has been imaged on QUEST at high temporal resolution (50 kHz) using a fast framing visible camera. Statistical features of fluctuations including blobs are investigated in both open and closed magnetic configurations. Higher order moments of fluctuations including blobs and their mutual relationships are investigated. Close resemblance of the density pdf with Gaussian or beta distributions has earlier been predicted. Hence, an attempt has been made to characterize the density pdf with respect to the Pearson system to ascertain the implications of the stochastic forces.

DOI： 10.1109/TPS.2011.2163423

記述言語：英語  

CCS (Cauchy Condition Surface) method is a numerical approach to reproduce plasma shape, which has good precision in conventional tokamaks. In order to apply it in plasma shape reproduction of ST (Spherical Tokamak). the calculation precision of the CCS method in a spherical tokamak CPD (Compact PWI experimental Device)(B(t) = 0.25 T, R = 0.3 m, a = 0.2 m) has been analyzed. The precision was confirmed also in ST and decided to be applied to a spherical tokamak QUEST (B(t), = 0.25 T, R = 0.68 m, a = 0.40 m).. In present stage from the magnetic measurement, it is known that the eddy current effect is large in QUEST experiment, and there are no special magnetic measurements for eddy current now, so some proper model should be selected to evaluate the eddy current effect. The eddy current density by not only CS (Center Solenoid) coil but also plasma current is calculated using EDDYCAL (JAEA). The eddy current magnitudes are taken as unknown variables and solved together with plasma shape reconstruction in ohmic discharge and ECCD (Electron Cyclotron Current Drive) discharge. (C) 2011 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2011.04.045

記述言語：英語  

Because QUEST tokamak has a divertor configuration with a higher κ and a negative n-index, a precise power supply with a rapid response is needed to control the vertical position of the plasma. A matrix converter is a direct power conversion device that uses an array of controlled bidirectional switches as the main power elements for creating a variable-output current system. This paper presents a novel three-phase to two-phase topological matrix converter as a proposed power supply that stabilizes the plasma vertical position and achieves unity input power factor. An indirect control strategy in which the matrix converter is split into a virtual rectifier stage and a virtual inverter stage is adopted. In the virtual rectifier stage, the instantaneous active power and reactive power are decoupled on the basis of system equations derived from the DQ transformation; hence, unity power factor is achieved. Space vector pulse width modulation is adopted to determine the switching time of each switch in the virtual rectifier; the output voltage of the virtual rectifier is adjusted by the virtual inverter stage to obtain the desired load current. Theoretical analyses and simulation results are provided to verify its feasibility.

記述言語：英語  

記述言語：英語  

The 16-element phased-array antenna has been developed for the EBWH/CD experiments. The 2-D k-spectra could be controlled. The heat load and stress were analyzed to be moderate in the 200kW CW operation. The symmetrically 2-D narrow-field distributions were excited by the developed antenna, indicating its good applicability to the EBWH/CD experiments.

DOI： 10.1109/irmmw-THz.2011.6105070

記述言語：英語  

The 16-element phased-array antenna has been developed for the EBWH/CD experiments. The 2-D k-spectra could be controlled. The heat load and stress were analyzed to be moderate in the 200kW CW operation. The symmetrically 2-D narrow-field distributions were excited by the developed antenna, indicating its good applicability to the EBWH/CD experiments.

DOI： 10.1109/irmmw-THz.2011.6105070

記述言語：英語  

記述言語：英語  

A study of radial propagation and electric fields induced by charge separation in blob-like structures has been performed in a non-confined cylindrical electron cyclotron resonance heating plasma on Q-shu University Experiment with a Steady-State Spherical Tokamak using a fast-speed camera and a Langmuir probe. The radial propagation of the blob-like structures is found to be driven by E x B drift. Moreover, these blob-like structures were found to have been accelerated, and the property of the measured radial velocities agrees with the previously proposed model [C. Theiler et al., Phys. Rev. Lett. 103, 065001 (2009)]. Although the dependence of the radial velocity on the connection length of the magnetic field appeared to be different, a plausible explanation based on enhanced short-circuiting of the current path can be proposed. (C) 2011 American Institute of Physics. [doi:10.1063/1.3640494]

DOI： 10.1063/1.3640494

記述言語：英語  

記述言語：英語  

記述言語：英語  

Blob generation mechanisms have been studied in slab plasma produced by electron cyclotron waves in a simple magnetic configuration with vertical (B-z) and toroidal fields (B-t). Two dimensional high speed images are analyzed as plasma and fluctuation evolution. Near the outer boundary (source region) of the slab plasma helix-sinusoidal perturbations are exited, whose helix angle and vertical wavelength are consistent with pitch angle and turn distance of the magnetic field lines there. Two dimensional evolution of these helix instabilities are studied as a function of B-z/B-t. Steep gradient outside the source region triggers the plasma to expand radially as a blob ejection. Along the ridge line of the maximum inverse scale length a wave front of the helix perturbation moves outward. Radial ejection of blob with typical velocity of similar to 1 km/s is recognized as non-linear evolution of these helix instabilities. Acceleration of a blob is found in the source-free region. (C) 2010 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2010.11.040

記述言語：英語  

Electron cyclotron resonance heating (ECRH)-assisted plasma breakdown is foreseen with full and half magnetic field in ITER. As reported earlier, the corresponding O1- and X2-schemes have been successfully used to assist pre-ionization and breakdown in present-day devices. This contribution reports on common experiments studying the effect of toroidal inclination of the ECR beam, which is >= 20 degrees in ITER. All devices could demonstrate successful breakdown assistance for this case also, although in some experiments the necessary power was almost a factor of 2 higher compared with perpendicular launch. Differences between the devices with regard to the required power and vertical field are discussed and analysed. In contrast to most of these experiments, ITER will build up loop voltage prior to the formation of the field null due to the strong shielding by the vessel. Possible consequences of this difference are discussed.

DOI： 10.1088/0029-5515/51/8/083031

記述言語：英語  

Blob-like structures are observed in an open magnetic field line configuration of electron cyclotron resonance (ECR) heating plasma in QUEST. The combined measurement with a fast camera and a movable Langmuir probe show that visible filamentary structures extended along field lines do correspond to over-dense plasma structures. Ion saturation current time trace shows a front-like asymmetric burst when a visible structure reaches the probe during its outward propagation. The typical radial velocity of the blob-like structure, V-b, was measured as similar to 1 km/s with the fast camera at the intermediate area. V-b increased as the blob-like structure moved along major radius, R (outward). The average velocity of E x B drift derived from the probe measurement, slightly larger than V-b, was also increased as the blob-like structure propagated to larger R. (C) 2010 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2010.12.225

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

The phased‐array antenna system for Electron Cyclotron/Bernstein Wave Heating and Current Drive experiments has been developed in the QUEST. The antenna was designed to excite a pure O‐mode wave in the oblique injection for the O‐X‐B mode conversion experiments, and its good performances were confirmed at a low power level. The plasma current (<∼15 kA) with an aspect ratio of 1.5 was started up and sustained by only RF injection in the low‐density operations. The long pulse discharge of 10 kA was also attained for 37 s. The new density window to sustain the plasma current was observed in the high‐density plasmas. The single‐null divertor configuration with the high plasma current (<∼25 kA) was attained in the 17 s plasma sustainment.

記述言語：日本語  

記述言語：英語  

Ohmic plasma currents of up to 17 kA with a discharge duration of 0.32 s have been obtained in the Kyushu University Experiment with Steady-State Spherical Tokamak (QUEST) with the help of electron cyclotron wave (ECW) and cancellation coils (CCs). The CCs, originally installed to create a field null in the plasma breakdown phase, are essential for producing plasma current in QUEST. Although the ohmic coil current is initially biased and then reduced completely to zero to induce the plasma current in 15-20 ms, we demonstrate that the flat top of the plasma current exceeding 20 ms is maintained by the vertical field after the ohmic current is switched off. This type of operation is quite favorable for extending pulsed operation to the steady state by electron Bernstein wave current drive (EBCD).

記述言語：英語  

記述言語：英語  

A permeation measuring system with a nickel membrane of 30 pm thickness was installed near the mid plane of the spherical tokamak, QUEST. Hydrogen permeation through the membrane heated at fix temperatures (422 -506 K) was measured during short pulse (< 1 s) and long pulse (1 hour) plasma discharges. After the membrane was heated to a required temperature, hydrogen plasma was discharged using a 2.45 GHz or 8.2 GHzRF system. Significant plasma-driven permeation was observed even for very short plasma discharges (e.g. 0.1s). Numerical calculations with the use of diffusion equation under recombination boundary conditions were conducted to simulate the transient permeation behavior. The numerical calculations were also used to estimate diffusion coefficient and recombination coefficients of membrane material. Temperature dependence of both the coefficients was explained by the Arrhenius law. A one hour long permeation curve was also numerically reproduced using the same set of parameters except an increasing recombination coefficient on the plasma side of the membrane.

DOI： 10.13182/FST11-A12719

記述言語：英語  

This paper presents a scheme for combining a Langmuir probe with a fast camera for measurement of the behavior of blob-like structures in the boundary region of an electron cyclotron resonance (ECR) heating plasma in Q-shu University Experiment with a Steady-State Spherical Tokamak (QUEST, major radius, R = 0.68 m, minor radius, a = 0.40 m, and toroidal magnetic field, Bt = 0.25 T at R = 0.64 m). The frame rate of the camera was typically set to 40,000 frames per second (FPS) with 192 × 144 pixels per frame. Radial motion of blob-like structures was observed in the half of the plasma space where the probe head was located. A radially movable and rotatable probe system was used to measure the floating potential from single unbiased tips, the potential of the positively biased tip, and the ion saturation current in two orthogonal directions in the outboard midplane region. Time series of the ion saturation current measured by the Langmuir probe and of pixels in a 40,000 FPS movie were compared and cross-correlated. The results of the two diagnostics agreed well, and the spatial scale was found to be of the same as the size of the probe head. The ion saturation current was asymmetric in terms of the time the blob-like structure was passing; fast camera imaging also clearly demonstrated the blobs' filamentary structures and radial motion at the edge of QUEST. This means that plasma in the blob-like structure hunches over, like blobs in other devices. The typical radial velocity of the structures is ∼1 km/s, and the structures were accelerated along their path of radial motion from the inner to the outer parts of the vacuum vessel.

記述言語：英語  

It is important to determine the best method for reconstructing the magnetic flux when eddy currents are significantly induced during magnetic measurement in spherical tokamaks (STs). Four methods for this reconstruction are investigated, and the calculated magnetic fluxes are compared to those measured in the cavity of a vacuum vessel. The results show that the best method is the one that uses currents from virtual coils for reconstruction. In this method, the placement of the virtual coils is optimized with numerical simulations using the Akaike information criterion (AIC), which indicates the goodness of fit of models used to fit measured data. The virtual coils are set on a line 15 cm outside the vacuum vessel.

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

The role of energetic electrons during the non-inductive current ramp-up is investigated in electron cyclotron heated plasma (≤ 60kW) in the spherical tokamak QUEST. The pulse height analysis of the hard X-ray (HX) is carried out with a fast time resolution of ≥1 ms, and acceleration of the electrons are studied by measuring the evolution of spectrum. It is found that wall recycling condition affects the formation of the tokamak configuration from an initial open magnetic field configuration. The vertical magnetic field Bz dependence of driven current Ip (~ 11 kA) is compared with those of HXs for studying contribution of fast electron. For high recycling case, Ip and HXs show peaks at an optimal low value of B_z/B_t = 7 ×10^-3, however, for low recycling case, both increase monotonically with B _z/B_t = 7 ×10^-3 - 21×10 ^-3.

記述言語：英語  

Particle retention and recycling in plasma fusion devices are generally associated with the diffusion of atomic hydrogen into the materials. The resulted permeation of atomic hydrogen is known as plasma driven permeation (POP). This permeation may also be significant, even in the walls, which are not directly exposed to the plasma. Under similar conditions, the permeation flux (Gamma(perm)) of hydrogen through a 30 mu m thick Ni membrane heated at 412-575K has been measured in the spherical tokamak QUEST. Gamma(perm) is being measured during the scans of different operating parameters like RF power (P(RF)), chamber pressure (P(chamber)), discharge widths (tau(dis)) and vertical magnetic field (B(Z)). Simultaneously edge plasma density and spectral intensities of atomic (Balmer) lines and molecular (Fulcher) bands have been compared with the permeation measurements. A linear relationship has been established between the time integrated Gamma(perm) i.e. permeation fluence (Q(perm)) and the time integrated H(alpha) intensity i.e. H(alpha) fluence (Q(alpha)). Q(perm) also shows a strong relationship with the edge plasma density and various spectral fluences. The obtained results are discussed for exploring the applicability of the permeation probes in measuring the atomic flux near the first walls. (C) 2010 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2010.08.032

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

The square waveguide phased-array antenna system has been developed for Electron Bernstein Emission diagnostics. The broadband orthomode transducer was prepared to measure the wave polarization. The measured fields were in good agreements with those evaluated from the Kirchhoff integral.

DOI： 10.1109/ICIMW.2010.5612710

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

The differential-phase reflectometry has been developed to measure the density profile, concerning electron Bernstein Wave (EBW) heating and current drive in the QUEST. The Phased Array Antenna (PAA) system was proposed for the reflectometry to improve the ratio of signal to noise. The PAA was confirmed to be effective, as a launching antenna to receive the larger reflected-wave signal.

DOI： 10.1109/ICIMW.2010.5612434

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

The CW phased-array antenna for the EBWH/CD has been developed in QUEST. The measured fields were in excellent agreement with those evaluated by the Kirchhoff integral. The thermal load and stress were analyzed. The phase array has been fast scanned to control the incident polarization and angle.

DOI： 10.1109/ICIMW.2010.5612352

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

The square waveguide phased-array antenna system has been developed for Electron Bernstein Emission diagnostics. The broadband orthomode transducer was prepared to measure the wave polarization. The measured fields were in good agreements with those evaluated from the Kirchhoff integral.

DOI： 10.1109/ICIMW.2010.5612710

記述言語：英語  

The CW phased-array antenna for the EBWH/CD has been developed in QUEST. The measured fields were in excellent agreement with those evaluated by the Kirchhoff integral. The thermal load and stress were analyzed. The phase array has been fast scanned to control the incident polarization and angle.

DOI： 10.1109/ICIMW.2010.5612352

記述言語：英語   掲載種別：研究論文（学術雑誌）  

The differential-phase reflectometry has been developed to measure the density profile evolution, concering electron Bernstein Wave (EBW) heating and current drive in the QUEST. The proper phase dependence was confirmed in developed single-side-band heterodyne differential-phase reflectometry at RIAM microwave/mm-wave test facilities. The phase time evolutions in the QUEST were also successfully measured, but the weak reflected-wave caused the small ratio of signal to noise in the measurements. The Phased Array Antenna (PAA) system was proposed for the reflectometry to obtain the large reflected-wave. The oblique viewing EBW emission can be measured by the precise phasearray measurements between the PAA waveguide elements. The PAA was confirmed to be effective, as a launching antenna for the reflectometry, and as a receiving antenna for the EBW radiometry. Keywords: AM reflectometry, Electron Bernstein wave, phased-array antenna, QUEST

記述言語：英語  

記述言語：英語  

The differential-phase reflectometry has been developed to measure the density profile evolution, concering electron Bernstein Wave (EBW) heating and current drive in the QUEST. The proper phase dependence was confirmed in developed single-side-band heterodyne differential-phase reflectometry at RIAM microwave/mm-wave test facilities. The phase time evolutions in the QUEST were also successfully measured, but the weak reflected-wave caused the small ratio of signal to noise in the measurements. The Phased Array Antenna (PAA) system was proposed for the reflectometry to obtain the large reflected-wave. The oblique viewing EBW emission can be measured by the precise phasearray measurements between the PAA waveguide elements. The PAA was confirmed to be effective, as a launching antenna for the reflectometry, and as a receiving antenna for the EBW radiometry. Keywords: AM reflectometry, Electron Bernstein wave, phased-array antenna, QUEST

DOI： 10.1109/ICIMW.2010.5612434

記述言語：英語  

The differential-phase reflectometry has been developed to measure the density profile evolution, concering electron Bernstein Wave (EBW) heating and current drive in the QUEST. The proper phase dependence was confirmed in developed single-side-band heterodyne differential-phase reflectometry at RIAM microwave/mm-wave test facilities. The phase time evolutions in the QUEST were also successfully measured, but the weak reflected-wave caused the small ratio of signal to noise in the measurements. The Phased Array Antenna (PAA) system was proposed for the reflectometry to obtain the large reflected-wave. The oblique viewing EBW emission can be measured by the precise phasearray measurements between the PAA waveguide elements. The PAA was confirmed to be effective, as a launching antenna for the reflectometry, and as a receiving antenna for the EBW radiometry. Keywords: AM reflectometry, Electron Bernstein wave, phased-array antenna, QUEST

DOI： 10.1109/ICIMW.2010.5612434

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

After the non-inductive current startup experimental campaign on the spherical tokamak compact plasma wall interaction device (CPD), various localized damage tracks or footprint traces have been observed on plasma facing components (PFCs), such as the chamber walls and the 'stiffeners' that support them. Although the magnetic field configuration is mainly open, the footprint traces are classified as (1) radially distributed traces, (2) toroidal imperfect circular traces with small gaps, (3) arc-shaped traces and (4) vertically distributed traces. The surface analysis of the samples attached near the traces has been carried out by scanning electron microscopy and x-ray photoelectron spectroscopy. They suggest thin deposition of impurity materials (C, Cu, Ti, Fe and their oxides) over the traces. These footprint traces are analysed in view of the localized plasma wall interaction and the loss of energetic electrons using orbit calculations. Radially distributed traces correspond to the loss of co- and countermoving passing electrons mainly escaping along the magnetic field lines. The imperfect circular traces are found corresponding to lost orbits of the energetic trapped electrons largely crossing the magnetic field lines. Other traces are also discussed from a viewpoint of loss along the magnetic field lines and impurity deposition.

DOI： 10.1088/0029-5515/50/2/025017

記述言語：英語  

記述言語：英語  

QUEST focuses on the steady state operation of the spherical tokamak by controlled PWI and electron Bernstein wave current drive. One of the main purposes of QUEST is an achievement of long duration discharge with MW-class injected power. As the result, QUEST should be operated in the challenging region on heat and particle handling. To do the particle handling, high temperature all metal wall up to 623K and closed divertors are planned, which is to realize the steady-state operation under recycling ratio, R = 1. This is a dispensable check to DEMO, because wall pumping should be avoided as possible in the view of tritium retention. The QUEST project will be developed in increment step such as, I. low β steady state operation in limiter configuration, II. low β steady state operation in divertor configuration, III. relatively high β steady state operation in closed divertor configuration. Phase I in the project corresponds to these two years, and final goal of phase I is to make full current drive plasma up to 20 kA. Closed divertor will be designed and tested in the Phase II. QUEST is running from Oct., 2008 and the first results are introduced.

DOI： 10.1585/pfr.5.S1007

記述言語：英語  

It is important to determine the best method for reconstructing the magnetic flux when eddy currents are significantly induced during magnetic measurement in spherical tokamaks (STs). Four methods for this reconstruction are investigated, and the calculated magnetic fluxes are compared to those measured in the cavity of a vacuum vessel. The results show that the best method is the one that uses currents from virtual coils for reconstruction. In this method, the placement of the virtual coils is optimized with numerical simulations using the Akaike information criterion (AIC), which indicates the goodness of fit of models used to fit measured data. The virtual coils are set on a line 15 cm outside the vacuum vessel.

DOI： 10.1585/pfr.5.S2083

記述言語：英語  

Various phenomena of plasma-wall interactions during long duration discharges in TRIAM-1M are investigated from macroscopic and microscopic viewpoints. It is found that the density dependence of the hydrogen neutral flux decay length is not very sensitive to the density (i.e. n_e^-0.2 ). Neutral transport through the scrape-off layer is important for structural formation of hydrogen recycling. The hydrogen retention in the co-deposited layer, which was obtained using a material probe during long duration discharges, is consistent with the global wall pumping rate estimated from particle balance analysis. The wall temperature and co-deposition play important roles on hydrogen re-emission and hydrogen absorption, respectively. Oxygen impurities should affect the erosion and deposition properties of the plasma-facing wall as well as the hydrogen retention property of the co-deposited layer. The co-deposition of hydrogen with molybdenum and the oxygen impurities both play important roles in the hydrogen recycling.

DOI： 10.1585/pfr.5.S2009

記述言語：英語  

This paper presents a scheme for combining a Langmuir probe with a fast camera for measurement of the behavior of blob-like structures in the boundary region of an electron cyclotron resonance (ECR) heating plasma in Q-shu University Experiment with a Steady-State Spherical Tokamak (QUEST, major radius, R = 0.68 m, minor radius, a = 0.40 m, and toroidal magnetic field, B_t = 0.25T at R = 0.64 m). The frame rate of the camera was typically set to 40,000 frames per second (FPS) with 192 ×144 pixels per frame. Radial motion of blob-like structures was observed in the half of the plasma space where the probe head was located. A radially movable and rotatable probe system was used to measure the floating potential from single unbiased tips, the potential of the positively biased tip, and the ion saturation current in two orthogonal directions in the outboard midplane region. Time series of the ion saturation current measured by the Langmuir probe and of pixels in a 40,000 FPS movie were compared and cross-correlated. The results of the two diagnostics agreed well, and the spatial scale was found to be of the same as the size of the probe head. The ion saturation current was asymmetric in terms of the time the blob-like structure was passing; fast camera imaging also clearly demonstrated the blobs' filamentary structures and radial motion at the edge of QUEST. This means that plasma in the blob-like structure hunches over, like blobs in other devices. The typical radial velocity of the structures is ~1km/s, and the structures were accelerated along their path of radial motion from the inner to the outer parts of the vacuum vessel.

DOI： 10.1585/pfr.5.S2077

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

The phased-array antenna has been designed for steady-state electron Bernstein wave heating and current drive at 8.2 GHz in the QUEST tokamak. The prototype antenna had rather large side-lobe components. In the new designed antenna the side-lobe components were reduced by modification of the antenna structure. For the steady-state operation, forced water-cooling and thermal stress were analyzed for the antenna component using the HFSS/ePhysics/ANSYS codes.

DOI： 10.1109/ICIMW.2009.5324704

記述言語：英語  

Two-dimensional structure of density fluctuations is examined during the current jump phase, indicating a change from the open magnetic fields to the closed ones. During the smooth current ramp-up phase the two-dimensional contour of the LiI intensity shows vertically alignment, consistent with the magnetic surfaces. At the inflection point in I_p ramp-up the LiI intensity contour becomes flat in the observation regime and then suddenly a steep gradient and higher intensity regime are formed in the vertical direction. This higher intensity corresponds to a burst of LiI waveform. According to these changes in the contour, it is found that, within 1 ms around the burst of LiI, a low frequency coherent wave with a long wavelength rapidly grows. The relations with other signals (magnetic flux and microwave stray power) are discussed with respect to the topological change in the magnetic configuration and mode conversion of the incident electromagnetic waves.

DOI： 10.1088/1009-0630/11/4/06

記述言語：英語  

In a spherical tokamak (ST), an operating frequency of the reflectometry is low due to the low magnetic field of the device. In the large ST, the reflected wave signal level is weak in the large propagating length, and a direct coupling between the launching and receiving antennae may become comparable level to the reflected signal due to the diffraction effect in the low operating frequency at the ST. New single-side-band heterodyne differential-phase reflectometry is proposed on a large ST device of the QUEST, to remove the direct coupling effect in an amplitude modulation reflectometry.

DOI： 10.1109/ICIMW.2009.5324785

記述言語：英語  

The QUEST (Kyushu, pronounced as "Q-shu" University Experiment with Steady-state spherical Tokamak) project is promoted from 2005 in the frame work of bi-directional collaboration organized by N1FS (National Institute for Fusion Science) to study plasma wall interaction on high temperature all-metal wall in long duration plasma on a spherical tokamak configuration. Present status and future plan of QUEST are presented.

DOI： 10.1541/ieejfms.129.589

記述言語：英語  

記述言語：英語  

記述言語：英語  

The phased-array antenna has been designed for steady-state electron Bernstein wave heating and current drive at 8.2 GHz in the QUEST tokamak. The prototype antenna had rather large side-lobe components. In the new designed antenna the side-lobe components were reduced by modification of the antenna structure. For the steady-state operation, forced water-cooling and thermal stress were analyzed for the antenna component using the HFSS/ePhysics/ANSYS codes.

DOI： 10.1109/ICIMW.2009.5324704

記述言語：日本語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

Two approaches associated with wall recycling have been performed in a small spherical tokamak device CPD (compact plasma wall interaction experimental device), that is, (1) demonstration of active particle recycling control, namely, 'active wall pumping' using a rotating poloidal limiter whose surface is continuously gettered by lithium and (2) a basic study of the key parameters which discriminates between 'wall pumping and fuelling'. For the former, active control of 'wall pumping' has been demonstrated during 50kW RF current drive discharges whose pulse length is typically similar to 300 ms. Although the rotating limiter is located at the outer board, as soon as the rotating drum is gettered with lithium, hydrogen recycling measured with H a spectroscopy decreases by about a factor of 3 not only near the limiter but also in the centre stack region. Also, the oxygen impurity level measured with O II spectroscopy is reduced by about a factor of 3. As a consequence of the reduced recycling and impurity level, RF driven current has nearly doubled at the same vertical magnetic field. For the latter, global plasma wall interaction with plasma facing components in the vessel is studied in a simple torus produced by electron cyclotron waves with I(p) < 1 kA. A static gas balance (pressure measurement) without external pumping systems has been performed to investigate the role of particle flux on a transition of 'wall fuelling' to 'wall pumping'. It is found that a critical particle flux exists to discriminate between them. Beyond the critical value, a large fraction (similar to 80&#37;) of pressure drop ('wall pumping') is found, suggesting that almost all injected particles are retained in the wall. Below it, a significant pressure rise ('wall fuelling') is found, which indicates that particles are fuelled from the wall during/just after the discharge. Shot history effects (integrated particle recycling behaviour from the plasma facing surfaces) are seen on that the critical particle flux is reducing from 0.8 x 10(-4) to similar to 0.1 x 10(-4) Torr during the experimental campaign (similar to 3000 shots). In the wall pumping pressure range the wall pumping fraction is reduced with increasing surface temperature up to 150 degrees C.

DOI： 10.1088/0029-5515/49/5/055020

記述言語：英語  

In a spherical tokamak (ST), an operating frequency of the reflectometry is low due to the low magnetic field of the device. In the large ST, the reflected wave signal level is weak in the large propagating length, and a direct coupling between the launching and receiving antennae may become comparable level to the reflected signal due to the diffraction effect in the low operating frequency at the ST. New single-side-band heterodyne differential-phase reflectometry is proposed on a large ST device of the QUEST, to remove the direct coupling effect in an amplitude modulation reflectometry.

DOI： 10.1109/ICIMW.2009.5324785

記述言語：英語  

記述言語：英語  

記述言語：英語  

Experiments to study wall recycling behaviour have been performed in the small spherical tokamak compact plasma-wall interaction experimental device (CPD) from the viewpoint of global as well as local plasma wall interaction condition. Electron cyclotron resonance (ECR) plasma of typically similar to 50 to 400 ms duration is produced using similar to 40 to 80 kW RF power. In order to study the global wall recycling behaviour, pressure measurements are carried out just before and after the ECR plasma in the absence of any external pumping. The recycling behaviour is found to change from release to pumping beyond a certain level of pressure value which is again found to be a function of shot history. The real-time local wall behaviour is studied in similar RF plasma using a rotating tungsten limiter, actively coated with lithium. Measurement of H. light intensity in front of the rotating surface has indicated a clear reduction (similar to 10&#37;) in the steady-state hydrogen recycling with continuous Li gettering of several minutes. (C) 2008 Elsevier BY. All rights reserved.

DOI： 10.1016/j.fusengdes.2008.05.009

記述言語：英語  

QUEST (R=0.68m, a=0.4m) focuses on the steady state operation of the spherical tokamak (ST) by controlled PWI and electron Bernstain wave (EBW) current drive (CD). The QUEST project will be developed along two phases, phase I: steady state operation with plasma current, Ip=20-30kA on open divertor configuration and phase II: steady state operation with Ip=100kA and β of 10&#37; in short pulse on closed divertor configuration. Feasibility of the missions on QUEST was investigated and the suitable machine size of QUEST was decided based on the physical view of plasma parameters. Electron Bernstein wave (EBW) current drive are planned to establish the maintenance of plasma current in steady state. Mode conversion efficiency to EBW was calculated and the conversion of 95&#37; will be expected. A new type antenna for QUEST has been fabricated to excite EBW effectively. The situation of heat and particle handling is challenging, and W and high temperature wall is adopted. The start-up scenario of plasma current was investigated based on the driven current by energetic electron and the most favorable magnetic configuration for start-up is proposed.

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

In TRIAM-1M, plasma discharge can be sustained for over five hours [H. Zushi, et al., Steady-state tokamak operation, ITB transition and sustainment and ECCD experiments in TRIAM-1M, Nucl. Fusion 45 (2005) S142-S156]. In order to avoid sitting in front of one console for the purpose of monitoring the plasma discharge, it is recommended that the experimental information be accessible from any location at any time. In addition, simple services to access experimental information are required in order to promote the participation of multiple researchers in the TRIAM-1M experiment. Thus, A WEB-based integrated data processing system that provides management for experiment planning, an experimental log, numerical data, and plasma supervision has been installed in the TRIAM-1M. These services are composed primarily of an Apache WEB server, a Tomcat JSP/Servlet container, and a MySQL relational database. This system is constructed using the object-oriented Java language, which is easy to maintain and develop because of the intrinsic characteristics of the Java language. When participating in experiments, researchers are required only to prepare a WEB browser on any platform and are no longer required to memorize complex operations because all services are provided with a uniform user interface through a WEB browser. Furthermore, with the integration of these services, the required information and numerical data can be provided promptly by tracing HTML links that are created dynamically by server applications. (C) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2007.12.012

記述言語：英語  

Superconducting tokamak TRIAM-1M has just ceased its operation in December 2005. However, in order to develop the systematic study on plasma-wall interaction in long duration discharges in the TRIAM-1M tokamak, spherical tokamak QUEST is under construction. To fulfill the mission of QUEST under a small group of the university in collaboration with other group, control and interlock system must be designed appropriately. Hence, a spherical tokamak CPD was constructed and the experiment is carried out for preparation of QUEST. The control and interlock system for QUEST is tested and developed during the experiment in the CPD. (c) 2008 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2008.01.004

記述言語：英語  

Fully non-inductive lower hybrid current drive (LHCD) plasmas have been achieved in TRIAM-1M and HT-7 superconducting tokamaks, but the current control in toroidal plasma was performed in different way during their steady-state operation (SSO). In HT-7, the plasma current was controlled mainly by the ohmic field, and partially by LHCD through ohmic transformer magnetic flux feedback control. In TRIAM-1M, the plasma current was controlled by lower hybrid wave (LHW) power directly according to a simple simulation model. (c) 2008 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2008.01.018

記述言語：英語  

On CPD [1] (Compact PWI experimental Device) experiment, information of electronic logbook and sequence status are distributed by Web services to prepare future experimental environment such as steady state operation and remote participation. Hence, all the researchers can acquire information with a Web browser installed on a personal computer if they are connected to the Internet. However, to carry a notebook computer all the time is a burden to researchers. Furthermore, the researchers may not be always connected to the Internet. Mobile phones are superior in portability compared to notebook computers, and are easy to connect with Internet through the wireless network of the telecom carriers. Moreover, since recent mobile phones have full browsing function, their affinities to the Web services are becoming high. On this account, Web services for mobile phones are developed to access experimental information. For sequence monitoring, a mobile application MIDlet that utilizes special functions of mobile phone such as sound and vibration is also developed to draw attentions of researchers to sequence status. (C) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2007.10.014

記述言語：英語  

記述言語：英語  

Two-dimensional lithium beam imaging technique has been applied in the spherical tokamak CPD (compact plasma wall interaction experimental device) to study the effects of magnetic field configurations on rf plasma boundary in the absence of any plasma current, and also for the measurement of a two-dimensional edge electron density profile. With the present working condition of the diagnostics, the minimum measured electron density can be similar to 1.0 X 10 16 m(-3); this is considered to be the definition for the plasma boundary. The performance of the lithium sheet beam is absolutely calibrated using a quartz crystal monitor. Experimental results reveal that magnetic field configuration, either mirror or so-called null, critically affects the rf plasma boundary. A sharp lower boundary is found to exist in magnetic null configuration, which is quite different from that in the weak mirror configuration. Theoretical calculations of particle drift orbit and. magnetic connection length (wall-to-wall) suggest that only mirror trapped particles are confined within a region where the magnetic connection length is similar to 4.0 in or more. A two-dimensional edge electron density profile is obtained from the observed Li I intensity profile. Overdense plasma formation is discussed from the viewpoint of mode conversion of rf wave into electron Bernstein wave and its dependence on the electron density profile. (c) 2008 American Institute of Physics.

DOI： 10.1063/1.2839034

記述言語：英語  

Two dimensional electron density profile measurement has been performed in the spherical tokamak CPD (Compact PWI experimental Device) using Li sheet beam imaging technique. The topological change from the open magnetic field line configuration to the closed one is visualized by this technique. The plasma current can be driven by RF waves itself in a weak mirror configuration and a clear change is observed in plasma boundary as well as magnetic field topology associated with the transition of the current from low (∼1 kA) to high (∼3 kA) value.

記述言語：英語  

On tokamak devices, an induction electric field induced by poloidal field (PF) coils plays a role to produce initial plasma. On a DIII-D tokamak, the required induction electric field for plasma breakdown agrees well with theoretical predictions based on the Townsend avalanche theory. According to the Townsend avalanche theory, the minimum induction electric field for plasma breakdown depends on neutral gas pressure and connection length. For stable plasma breakdown, a sufficiently large induction electric field is required. However, in the case of spherical tokamaks without electric insulation in the toroidal direction, the effect of eddy currents flowing in the toroidal direction should be considered in evaluating a feasible induction electric field because this effect suppresses the generation of an induction electric field. On a QUEST spherical tokamak, the possibility of Townsend avalanche breakdown is studied by evaluating the connection length and achievable induction electric field. The connection length is greater than 100 in in the case where a null point is set to be R = 0.56 in with a CS coil current of 2.0 kA and a PF26 coil current of 0.4 kA. Moreover, the induction electric field is about 1.5 V at this point including the effect of eddy currents. With these values, the initial plasma production by the induction electric field is sufficiently possible on QUEST.

DOI： 10.1143/JJAP.47.287

記述言語：英語  

Two dimensional electron density profile measurement has been performed in the spherical tokamak CPD (Compact PWI experimental Device) using Li sheet beam imaging technique. The topological change from the open magnetic field line configuration to the closed one is visualized by this technique. The plasma current can be driven by RF waves itself in a weak mirror configuration and a clear change is observed in plasma boundary as well as magnetic field topology associated with the transition of the current from low (~1 kA) to high (~3kA) value.

DOI： 10.1585/pfr.3.010

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

The amplitude modulation reflectometer system has been developed to measure the density profile in detail on the Compact PWI ( plasma-wall interaction ) Device (CPD) experiments. The probing frequency range was 6-18 GHz, and its frequency-switching time was 10 μs in an arbitrary step. Using the broadband fast-scanning reflectometer, the wide-range density profile can be measured with high time resolution. In X-mode reflectometory, the developed system successfully measured the phase delay of propagating waves at 6.6, 8.8 and 13.2 GHz, reflected at the vessel wall of the CPD and the R-cutoff layers in the CPD plasma.

DOI： 10.1109/icimw.2007.4516610

記述言語：英語  

The amplitude modulation reflectometer system has been developed to measure the density profile in detail on the Compact PWI ( plasma-wall interaction ) Device (CPD) experiments. The probing frequency range was 6-18 GHz, and its frequency-switching time was 10 μs in an arbitrary step. Using the broadband fast-scanning reflectometer, the wide-range density profile can be measured with high time resolution. In X-mode reflectometory, the developed system successfully measured the phase delay of propagating waves at 6.6, 8.8 and 13.2 GHz, reflected at the vessel wall of the CPD and the R-cutoff layers in the CPD plasma.

DOI： 10.1109/icimw.2007.4516610

記述言語：英語  

The amplitude modulation reflectometer system has been developed to measure the density profile in detail on the Compact PWI ( plasma-wall interaction ) Device (CPD) experiments. The probing frequency range was 6-18 GHz, and its frequency-switching time was 10 μs in an arbitrary step. Using the broadband fast-scanning reflectometer, the wide-range density profile can be measured with high time resolution. In X-mode reflectometory, the developed system successfully measured the phase delay of propagating waves at 6.6, 8.8 and 13.2 GHz, reflected at the vessel wall of the CPD and the R-cutoff layers in the CPD plasma.

DOI： 10.1109/icimw.2007.4516610

記述言語：英語  

記述言語：英語  

Fast cameras are powerful tool to visualize the edge turbulence in peripheral plasmas. Under Bi-directional collaborations recently in Heliotron J and GAMMA10 filamentary structures along the magnetic field line were firstly observed with GPI (gas puff imaging) by fast camera. In both machines the filamentary structures had similar stripe pattern in the images and simultaneous measurements of the ion saturation current by electrostatic probes show that the filamentary structures were relatively higher electron density/temperature regions in peripheral plasmas. It is not sufficient to conclude both filamentary structures are the same, however, these phenomena were thought to be related to the energy confinement. Thus the physics mechanism should be solved in the near future.

記述言語：英語  

記述言語：英語  

The amplitude modulation reflectometer system has been developed to measure the density profile in detail on the Compact PWI ( plasma-wall interaction) Device (CPD) experiments. The probing frequency range was 6-18 GHz, and its frequency-switching time was 10 mu s in an arbitrary step. Using the broadband fast-scanning reflectometer, the wide-range density profile can be measured with high time resolution. In X-mode reflectometory, the developed system successfully measured the phase delay of propagating waves at 6.6, 8.8 and 13.2 GHz, reflected at the vessel wall of the CPD and the R-cutoff layers in the CPD plasma.

DOI： 10.1109/icimw.2007.4516610

記述言語：英語  

Retention of hydrogen in co-deposits formed under high-density plasma discharge in TRIAM-1 M was studied. In order to quantify the retained hydrogen, material probe experiments were performed under the high-density ((n) over tilde (e) similar to 10(19) m(-3)) discharges. After the exposure to the plasma, the quantitative analysis of deposition, hydrogen retention, and microscopic modification of specimens were performed by means of ion beam analysis and transmission electron microscopy. The co-deposits mainly consisted of Mo. The deposition rate of Mo was about ten times higher than that of the low-density discharge case. The hydrogen concentrations (H/Mo) retained in the co-deposits were 0.06-0.17, which was much higher than that in bulk-Mo and almost equal to the low-density case. These results indicate that as long as the co-deposition layers are continuously formed.. strong wall pumping in TRIAM-1M is maintained during the discharges. (C) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2007.04.010

記述言語：英語  

記述言語：英語  

In order to investigate the surface temperature effects on plasma fuel recycling and impurity release from the plasma facing components, plasma discharges have been performed under selected plasma-wall interaction (PWI) conditions in the high-field superconducting tokamak, TRIAM-1M. By moving a water-cooled molybdenum movable limiter (ML) beyond the last closed flux surface, as defined by poloidal limiters, the surface temperature profile on it is varied. Hot spots have been observed on the ML surface in such conditions. The release behaviour of fuel as well as impurity particles from the ML surface has been studied as a function of hot spot temperature (T-hot) by means of wide range spectroscopy (200-1600 nm). A critical Thot is found to be similar to 2100 K above which the emission of both hydrogen and impurity particles enhances significantly. This is indicative of some thermally activated process playing an important role in PWIs between the limiter and the edge plasma. With the rise in hot spot temperature localized PWI at the ML is found to dominate the global recycling even when external fuelling is stopped.

DOI： 10.1088/0029-5515/47/8/018

記述言語：英語  

For initial plasma production, an induction electric field generated by applying voltage to a poloidal field (PF) coil system is used to produce a Townsend avalanche breakdown. When the avalanche margins are small, as for the International Thermonuclear Experimental Reactor (ITER) in which the induction electric field is about 0.3 V/m, the assistance of radio-frequency waves (RF) is provided to reduce the induction electric field required for reliable breakdown. However, the conditions of RF-assisted breakdown are not clear. Here, the effects of both RF and induction electric field on the RF-assisted breakdown are evaluated considering the electron loss. When traveling loss is the dominant loss, a simple model of an extended Townsend avalanche is proposed. In this model, the induction electric field required for RF-assisted breakdown can be decreased to half that required for induction breakdown.

DOI： 10.1143/JPSJ.76.084501

記述言語：英語  

Calorimetric measurements using the temperature increment of cooling-water were carried out to estimate the heat load distribution on the plasma facing components (PFCs) in the limiter discharges on TRIAM-1M. Line averaged electron density, n(e), and LH power, P-LH, dependences of the heat load on PFCs were measured. The heat load on the limiters was proportional to n(e)(1.5) in the range of n(e) = 0.2-1.0 x 10(19) m(-3) and P-LH(I) in the range of P-LH = 0.005-0.09 MW. For P-LH > 0.1 MW, the plasma transition to an enhanced current drive (ECD) mode appeared and the n(e) dependences on the heat load on the limiter moderated. This indicates that the heat flux to scrape-off layer (SOL) region was reduced due to the improvement of the plasma confinement. The up-down asymmetry of the heat load on the vacuum vessel was enhanced in the ECD mode, which may be caused by the increasing of the direct loss of energetic electrons. (c) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2007.01.255

記述言語：英語  

A sheet thermal Li beam to study the density contour near X-point and divertor strike points has been developed for the spherical tokamak CPD. The CCD camera with a time resolution of 1 ms and spatial resolution of similar to 1 mm can record two-dimensional images of LiI resonance line (300 mm x 500 mm). The performance of the sheet beam is absolutely calibrated by a quartz crystal monitor. The Li flux of 10(18)-10(19) m(-2) s(-1) is achieved near the X-point. The sheet characteristics of beam width and uniformity are also confirmed. This technique is applied during plasma initiation and electron cyclotron heating phase of plasma in a simple torus with a small vertical field. The formation of vertically stretched plasma at the position of the cyclotron resonance is clearly found. The expansion process towards the low field side is observed and an over dense plasma (> 1 x 10(18) m(-3)) is obtained. A model calculation of LiI emission with density range of 1-5 x 10(18) m(-3) agrees fairly well with the observations. Over dense plasma formation is also discussed. (c) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2007.01.259

記述言語：英語  

The observation of dusts in plasmas was carried out by high speed camera in full non-inductive lower hybrid current drive (LHCD) plasma on TRIAM-1M. The velocities of dusts were 10-50 m/s. The number of dust generated per second increased with the discharge duration in the range of 20-57 s. This suggests that a part of dusts were generated from a movable limiter whose the surface temperature increases with the discharge duration. Dusts were coming from various directions even close to the movable limiter. Dusts were collected in the vacuum vessel by use of a kind of cleaner and the composition and the size of dust were examined by Scanning Electron Microscope (SEM). Dust size was several mu m and composition was Molybdenum mixed with small amount of elements of stainless metal. The poloidal distribution of the collected mass of dusts was almost uniform. (c) 2007 Published by Elsevier B.V.

DOI： 10.1016/j.jnucmat.2007.01.199

記述言語：英語  

Tokamak operation has been performed under the localized plasma wall interaction conditions in TRIAM-1M, using a water cooled rail movable limiter (ML) made of Molybdenum. The limiter surface temperature dependence on hydrogen recycling and neutral molybdenum impurity production behaviour has been studied using visible as well as infrared spectroscopy during low power long pulse and high power short pulse discharges. In these experiments it is considered that escaped energetic electrons are lost mainly on ML and localised heat deposition makes a 'hot spot' on ML. It has been observed that H-2, H-alpha and MoI intensities critically depend on the hot spot temperature (T-hot) and enhanced recycling due to localized PWI on ML can fuel the discharge without any external fueling source. (c) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2007.01.117

記述言語：英語  

A diagnostic technique to study the global structure of the recycling and impurity (molybdenum) influx from the plasma facing components is proposed and is tested using the three kinds of perturbations (gas puffing, transport oscillation, and localized heat deposition). Balmer line intensities I-H alpha, charge exchange flux Gamma(CX) and neutral molybdenum line I-Mol are measured and their correlations are analyzed with response to the perturbation parameters (I-H alpha at the gas port, density n(e), and hot spot temperature T-hot). A simple model calculation is done to understand these correlations. It is found that the phase reversal of Gamma(CX) with respect to n(e) modulation is well reproduced and a critical density exists for the phase reversal of I-Mol. The evaporation Mo flux evaluated with measured T-hot is also compared with enhanced I-Mol for the heat load perturbation and it contribution to the total content of Mo ions is evaluated by similar to 30&#37; increment. (c) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2007.01.201

記述言語：英語  

An in situ and real time measurement system of erosion and deposition has been developed, which is based on interference of light on a thin semi-transparent layer of redeposited material on substrate. It has been applied to long duration discharges in TRIAM-1M. The sapphire window is used as substrate. The deposition pattern on the window indicates up down and toroidal asymmetry. In the 5 h 16 min discharge, the thickness of the deposited layer increased monotonically with time and its deposition rate is similar to 1.5 x 10(16) Mo m(-2) S-1. In the low density and long duration discharge, the Mo deposition rate on the window depends on the ratio of Mo flux to hydrogen flux. (c) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2007.01.055

記述言語：英語  

Oxygen impurity behaviors of one long duration discharge and repetition of short duration discharges have been investigated in TRIAM-1M. In the former case, the OII line intensity divided by the line averaged electron density, which is considered as a monitor of oxygen concentration on the plasma facing surface (PFS), decreased with the time constant, tau(d), of 30-50 s during the discharge due to the hydrogen flux to PFS. In the latter case, tau(d) is in the range of 70-600 s. There exists a big difference of global behavior of oxygen impurity between both cases. The difference seems to result from the absence or presence of the interval time between the discharges. The oxygen concentration on PFS increases during the interval time due to adsorption of H2O. The time constant of the increase in the oxygen concentration is evaluated to be about 5500 s from Langmuir adsorption isotherms analysis. (c) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2007.01.238

記述言語：英語  

記述言語：英語  

記述言語：日本語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

A simple model of radio frequency wave (RF) assisted breakdown based on Townsend avalanche theory is proposed for the purpose of evaluating the effect of RF on Townsend avalanche breakdown. According to this model, the required minimum electric field for RF-assisted breakdown can be decreased down to half of that for breakdown by the induction electric field alone. The electric field of RF reaches a minimum when the frequency of the RF is equal to the electron cyclotron frequency.

DOI： 10.1585/pfr.2.007

記述言語：英語  

Plasma shape reconstruction is important for plasma control of tokamak. Cauchy-Condition Surface (CCS) method is a numerical approach to reproduce plasma shape which has good precision in conventional tokamak. In order to apply it in the plasma shape reproduction of Compact PWI experimental Device (CPD) which is a new spherical tokamak in Kyushu University, the calculation precision of CCS method in CPD is analyzed in the paper.

DOI： 10.1585/pfr.2.S1095

記述言語：英語  

Combined experiments with lower hybrid (LH) waves (LHWs) and electron cyclotron (EC) waves (ECWs) have been performed to study counter (ctr) current drive (CD) (ctr-CD) in LHCD plasma. Although there exists a large spectrum gap under the condition of the forward (FW) LHCD at high ratio (>13) of the thermal velocity and the phase velocity, the relativistic Doppler resonance for backward (BW) ECWs coupled with the energetic electrons streaming along the current direction is used to confirm that the gap in the opposite velocity region is also filled by BW LHWs and the counter current tends to be driven. Three experimental scenarios have been studied for ctr-CD by the following experiments: (a) BW-LHW injection into FW LHW, (b) BW-ECW injection into FW LHW, and (c) BW-ECW injection into bidirectional BW-LHCD and FW-LHCD plasma. A transition depending on the power ratio is observed in case (a). The Ohkawa current is discussed for the co-driven current observation for case (b). The role of the amount of the resonance electrons is understood as a function of the power ratio of BW LHW to FW LHW for case (c).

DOI： 10.13182/FST07-A1503

記述言語：英語  

A new technique to measure the current profile in plasmas with asymmetric distribution function such as lower hybrid current drive (LHCD) by using hard X-ray (HXR) energy spectrum measurement along the top and bottom identical line of sights (ILOS) is proposed and is applied to the full and partially LHCD plasmas on the TRIAM-1M tokamak at the first time in the world. The pitch angles were measured at R - R
₀
= ±2.5cm, where R, R
₀
mean the major radii of the ILOS and the magnetic axis, respectively. The measured pitch angle of the magnetic field inverted at the magnetic axis estimated magnetic measurement in partially LHCD plasmas. This indicates that the difference of the measured pitch angles is caused by the plasma current in the plasma and this new method is available in detecting the current profile in tokamaks. In full LHCD plasma, no difference between the HXR signals along the top and bottomILOS appear. This indicates that the current density around the magnetic axis was reduced compared with that in partially LHCD plasmas. This observation is no contradiction with power deposition of LHCD.

DOI： 10.1585/pfr.2.S1071

記述言語：英語  

Fast cameras are powerful tool to visualize the edge turbulence in peripheral plasmas. Under Bi-directional collaborations recently in Heliotron J and GAMMA10 filamentary structures along the magnetic field line were firstly observed with GPI (gas puff imaging) by fast camera. In both machines the filamentary structures had similar stripe pattern in the images and simultaneous measurements of the ion saturation current by electrostatic probes show that the filamentary structures were relatively higher electron density/temperature regions in peripheral plasmas. It is not sufficient to conclude both filamentary structures are the same, however, these phenomena were thought to be related to the energy confinement. Thus the physics mechanism should be solved in the near future.

DOI： 10.1585/pfr.2.S1055

記述言語：英語  

記述言語：英語  

The limiter surface temperature dependence on hydrogen recycling and molybdenum impurity production behavior was studied with optical spectroscopy during long pulse plasma discharges in TRIAM-1M. It has been observed that the H
_α
intensity critically depends on the limiter surface temperature. The Mo-I intensity from the hot spot has shown a negative temperature dependence. This is believed to be due to the reduction of heat load and enhanced CX flux due to ion temperature rise.

DOI： 10.1109/FUSION.2005.252980

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

Internal Reconnection Event (IRE) is a characteristic relaxation phenomenon in STs (Sperical Tokamaks). To investigate the deformation of the plasma shape during IRE, four PIN diode arrays of 20 channels were installed on the TST-2@K (TST-2 at Kyushu University). Precursor of IRE was observed for several milli-seconds. The fluctuation was composed of two dominant components in frequency of 10kHz and 4kHz. The mode structure of 10kHz component is n/m=1/1 helical structure and 4kHz is n/m= 3/4. The overlap of modes (10kHz and 4kHz) was considered to be cause of IRE in TST-2 from the position and the growth of the modes.

記述言語：英語  

A new diagnostic technique to study the global structure of the recycling among the plasma facing components PFCs has been proposed and tested using the three kinds of perturbations (gas puffing, transport oscillation, and localized heat deposition) at the frequency less than a few Hz . H
_α
intensities are measured at various toroidal and poloidal positions and FFT analysis shows that the propagation velocity of the neutral hydrogen in the toroidal direction is < 25 m/s for the gas puff modulation. For toroidally uniform transport oscillations, on the other hand, in-out asymmetry of recycling structure is found. The localized heat deposition causes the localized particle source distribution in the torus due to enhanced efflux of the hydrogen particles only fueled from the hot area.

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

A new technique to measure the current profile in lower hybrid current drive (LHCD) plasmas by using hard X-ray (HXR) energy spectrum measurement along the top and bottom identical line of sight (ILOS) is proposed. An angle between ILOS and the magnetic filed line is out of alignment from the perpendicular by the poloidal magnetic field. As the emission cross-section of HXR strongly depends on the angle to the drift direction of energetic electrons, that is the magnetic filed line, this disagreement of the angles makes significant difference between the signals of detectors located on the top and bottom. As the result, the poloidal magnetic field is able to be derived from the difference of the HXR emission along the top and bottom ILOS. The method was applied to the full LHCD plasmas on the TRIMA-1M tokamak and the plasma current around the magnetic axis was estimated.

記述言語：英語  

記述言語：英語  

記述言語：英語  

In order to investigate the spectrum gap problem in LHCD and controllability for the current profile by means of counter current drive in full current drive plasma, experiments by combination of backward (BW) propagating LHW and/or ECW have been performed in a plasma sustained by forward (FW) propagating LHW. Three ctr-CD aspects have been investigated for relativistic resonance with respect to the power ratio of backward (PBW) to forward (PFW) LHWs; 1) a clear reduction of co-ICD for PBW/PFW < 0.8; 2) a rapid positive change in co-ICD and broadening in j(r) for PBW/PFW > 0.8 [H. Zushi et al., Nucl. Fusion 41 (2001) 1483]; and 3) a large positive change in co-ICD by ctr ECW into FW LHCD plasma and further changes to negative value in co-ΔICD depending on PBW/PFW of LHW.

記述言語：英語  

記述言語：英語  

記述言語：英語  

Several techniques for initiating the plasma current without the use of the central solenoid are being developed in TST-2. While TST-2 was temporarily located at Kyushu University, two types of start-up scenarios were demonstrated. (1) A plasma current of 4 kA was generated and sustained for 0.28 s by either electron cyclotron wave or electron Bernstein wave, without induction. (2) A plasma current of 10 kA was obtained transiently by induction using only outboard poloidal field coils. In the second scenario, it is important to supply sufficient power for ionization (100 kW of EC power was sufficient in this case), since the vertical field during start-up is not adequate to maintain plasma equilibrium. In addition, electron heating experiments using the X-B mode conversion scenario were performed, and a heating efficiency of 60% was observed at a 100 kW RF power level. TST-2 is now located at the Kashiwa Campus of the University of Tokyo. Significant upgrades were made in both magnetic coil power supplies and RF systems, and plasma experiments have restarted. RF power of up to 400 kW is available in the high-harmonic fast wave frequency range around 20 MHz. Four 200 MHz transmitters are now being prepared for plasma current start-up experiments using RF power in the lower-hybrid frequency range. Preparations are in progress for a new plasma merging experiment (UTST) aimed at the formation and sustainment of ultra-high beta ST plasmas.

DOI： 10.1088/0029-5515/46/8/S05

記述言語：英語  

記述言語：英語  

Plasma start-up and sustainment without an inductive field have been studied in the TST-2@K spherical tokamak using high power RF sources (8.2 GHz/up to 170 kW). Steady state discharges with a plasma current of 4 kA were achieved. The line integrated density was about 3 x 10(17) m(-2) and the electron temperature was 160 eV. A truncated equilibrium was introduced to reproduce magnetic measurements. It was found that a positive Pfirsch-Schluter current in the open field line region at the outboard boundary makes a significant contribution to the current. Insensitivity of the current to variations in the vertical field and RF power variation was also found.

DOI： 10.1088/0029-5515/46/7/003

記述言語：英語  

記述言語：英語  

A new technique to measure the current profile in lower hybrid current drive (LHCD) plasmas by using hard X-ray (HXR) energy spectrum measurement along the top and bottom identical line of sight (ILOS) is proposed. An angle between ILOS and the magnetic filed line is out of alignment from the perpendicular by the poloidal magnetic field. As the emission cross-section of HXR strongly depends on the angle to the drift direction of energetic electrons, that is the magnetic filed line, this disagreement of the angles makes significant difference between the signals of detectors located on the top and bottom. As the result, the poloidal magnetic field is able to be derived from the difference of the HXR emission along the top and bottom ILOS. The method was applied to the full LHCD plasmas on the TRIMA-1M tokamak and the plasma current around the magnetic axis was estimated.

記述言語：英語  

Internal Reconnection Event (IRE) is a characteristic relaxation phenomenon in STs (Sperical Tokamaks). To investigate the deformation of the plasma shape during IRE, four PIN diode arrays of 20 channels were installed on the TST-2@K (TST-2 at Kyushu University). Precursor of IRE was observed for several milli-seconds. The fluctuation was composed of two dominant components in frequency of 10kHz and 4kHz. The mode structure of 10kHz component is n/m=1/1 helical structure and 4kHz is n/m= 3/4. The overlap of modes (10kHz and 4kHz) was considered to be cause of IRE in TST-2 from the position and the growth of the modes.

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

A remote steering antenna was recently developed for electron cyclotron heating and current drive (ECH/ECCD) experiments on the TRIAM- I M tokamak. This is the first application of the remote steering antenna concept for ECH/ECCD experiments, which have conditions relevant to the International Thermonuclear Experimental Reactor (ITER). Fundamental ECH and ECCD experiments were conducted in the ITER frequency from the low field using this antenna system. In addition to the angles near 0 degrees, the launcher was a symmetric direction antenna with an extended steering-angle capability of +/- (8 degrees-19 degrees). The output beam from the antenna was a well-defined Gaussian with a proper steering angle. The Gaussian content and the steering-angle accuracy were 0.85 and -0.5 degrees, respectively. The high power tests measured the antenna transmission efficiency at 0.90-0.94. The efficiencies obtained in the low and high power tests were consistent with the calculations using higher-order modes. In order to excite the pure O/X-modes in the oblique injection, two polarizers were used to control the elliptical polarization of the incident beam for the ECCD experiments. The fundamental O/X-mode ECH/ECCD was applied to lower hyrid current drive plasmas at the optimized incident polarization. In the X-mode experiment, at medium density (similar to 1 x 10(19) m(-3)), clear differences in the plasma current and the hard x-ray intensity were observed between the co- and counter-steering injections due to the ECCD effect on the coupling of forward fast electrons.

DOI： 10.1088/0029-5515/46/5/001

記述言語：英語  

Intensity and phase profiles of the output beams from a remote steering antenna were measured in the anti-symmetric and symmetric directions. The beam properties are discussed with the imaging property calculations.

記述言語：英語  

The limiter surface temperature dependence on hydrogen recycling and molybdenum impurity production behavior was studied with optical spectroscopy during long pulse plasma discharges in TRIAM-1M. It has been observed that the H-alpha intensity critically depends on the limiter surface temperature. The Mo-I intensity from the hot spot has shown a negative temperature dependence. This is believed to be due to the reduction of heat load and enhanced CX flux due to ion temperature rise.

DOI： 10.1109/FUSION.2005.252980

記述言語：英語  

Conventional data acquisition systems cannot deal with data acquisition for a long-time discharge of a nuclear fusion reactor. Thus, continuous data acquisition with a real-time data presentation during discharge must be developed. Two data acquisition systems, which include alternating CAMAC data acquisition and long-time PCI data acquisition, are designed for the longtime operation of HT-7 tokamak. Since an effective alternating mode is adopted, the alternating CAMAC data acquisition can accurately and continuously acquire data at a rate of 10 kHz. The acquired data is immediately transmitted to a data server and real-time results can be presented during the plasma discharge. As for the long-time PCI data acquisition, a special kind of PCI A/D card, which has a hard disk on board, is designed to collect data at a max speed of 200 kHz. Thus, the total sampling duration is only related to the capacity of the hard disk on board. These two types of data acquisitions were applied to HT-7 tokamak and a 250 s discharge was acquired. These data acquisition systems were also successfully demonstrated on a 2500 s plasma discharge on TRIAM-1M. This paper describes the two data acquisitions in detail. (c) 2005 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2005.09.055

記述言語：英語  

Magnetic sensorless control experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. Previously the horizontal position was calculated from the vertical field coil current and voltage without using signals of magnetic sensors like magnetic coils and flux loops placed near the plasma. The calculations are made focusing on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem of integrator of magnetic sensors. Two kinds of experiments were carried out, to keep the position constant and swing the position in a triangular waveform. (c) 2005 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2005.08.094

記述言語：英語  

Magnetic sensorless sensing and control experiments with the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The sensing is made to focus on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem with the integrator of magnetic sensors. Two kinds of control experiments have been carried out: to keep the position constant and swing the position in a triangular waveform. And magnetic sensorless sensing of plasma shape is discussed.

DOI： 10.1088/1009-0630/8/1/18

記述言語：英語   掲載種別：研究論文（その他学術会議資料等）  

In TRIAM-1M, the influence of the local structure of the first wall on the global wall recycling has been studied using the ML with good cooling capability. No wall saturation was observed in the 5 h 16 min discharge with the ML, since particle release was suppressed by less temperature rise of the main chamber. The global wall pumping rate without the ML, ~8.6 x 10
¹⁶
atoms m
^-2
s
^-1
, is 3.6 times higher than that of the long duration discharge without the ML. The temperature rise in the wall is attributed to the transition of the wall role from the particle sink to the source. The co-deposition of hydrogen with Mo is a key for the wall role of the particle sink. No clear difference of the growth rate of Mo deposits on the viewing window located from the limiter surface between with and without the ML. The growth rate is from ~2 x 10
^-4
nm/s to ~8 x 10
^-4
nm/s. The toroidal structure of hydrogen recycling changes significantly near at the ML. However, integration of H intensity with and without the ML over the whole toroidal length is about the same within 3% , which is less than the error, i.e. 15%.

記述言語：英語  

Experiments aiming at 'day long operation at high performance' have been carried out in TRIAM-1M. The record value of the discharge duration was updated to 5 h and 16 min. Steady-state tokamak operation is studied under the localized plasma wall interaction conditions. The distributions of the heat load, the particle recycling flux and impurity source are investigated to understand the co-deposition and wall pumping. The formation and sustainment of an internal transport barrier (ITB) in enhanced current drive mode has been investigated by controlling the lower hybrid driven current profile by changing the phase spectrum. An ITER relevant remote steering antenna for electron cyclotron wave injection was installed and a relativistic Doppler resonance of the oblique propagating extraordinary wave with energetic electrons driven by lower hybrid waves was studied.

DOI： 10.1088/0029-5515/45/10/S12

記述言語：英語  

Plasma current (I_p) initiation and ramp-up utilizing induction by poloidal field (PF) coils located on the outboard side of the torus were demonstrated in both conventional aspect ratio tokamak (100 kA in JT-60U) and spherical tokamak (10 kA in TST-2). In the presence of sufficient source of plasma by RF ionization (1 MW in JT-60U, 100 kW in TST-2), I_p formation is possible without the existence of a PF null, with the initial vertical field (B_v) in the direction opposite to that required for equilibrium. Utilization of the extra flux swing provided by the PF coils before B_v switches sign is important in arriving at a high enough I _p that matches B_v at the end of PF coil current ramp. In JT-60U, recharging of the OH coil was observed with only perpendicular and counter NB injection, suggesting overdrive by bootstrap current. In TST-2, a quasi-steady-state I_p of 4 kA was formed and sustained by RF power (P_RF) alone, without the use of induction. The plasma current centroid of this type of plasma is located on the outboard side. The dependences of I_p on P_RF and B_v were found to be weak. The results presented here offer encouragement to CS-less operation of a tokamak fusion reactor, which should have a large impact on the economic competitiveness.

DOI： 10.1063/1.2098259

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

In TRIAM-1M, the influence of the local structure of the first wall on the global wall recycling has been studied using the ML with good cooling capability. No wall saturation was observed in the 5 h 16 min discharge with the ML, since particle release was suppressed by less temperature rise of the main chamber. The global wall pumping rate without the ML, ~8.6 x 10
¹⁶
atoms m
^-2
s
^-1
, is 3.6 times higher than that of the long duration discharge without the ML. The temperature rise in the wall is attributed to the transition of the wall role from the particle sink to the source. The co-deposition of hydrogen with Mo is a key for the wall role of the particle sink. No clear difference of the growth rate of Mo deposits on the viewing window located from the limiter surface between with and without the ML. The growth rate is from ~2 x 10
^-4
nm/s to ~8 x 10
^-4
nm/s. The toroidal structure of hydrogen recycling changes significantly near at the ML. However, integration of H intensity with and without the ML over the whole toroidal length is about the same within 3% , which is less than the error, i.e. 15%.

記述言語：英語  

記述言語：英語  

A remote steering antenna designed as a symmetric direction antenna is prepared for Electron Cyclotron Heating (ECH) and Current Drive (ECCD) experiments on the TRIAM-1M tokamak. The steering angles of 8-19 degrees are available in the design, in addition to the angle near 0 degrees. The antenna is tested at both of high and low power levels. The fraction of radiated power to injected power at the antenna is roughly evaluated at the low power test. The angle radiated from the antenna is checked from both of the intensity and phase measurements at the low power test. Several percentage losses at the antenna are evaluated at the high power test. The antenna system is evaluated to work correctly.

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

Internal transport barrier (ITB) has been obtained in full lower hybrid current driven (LHCD) plasmas on a superconducting tokamak, TRIMA-1M (R=0.84m, a x b=0.12mx0.18m, BT<8T). The formation of ITB depends on the current density profile, j(r), varied by the power deposition of the lower hybrid (LH). The plasma with ITB can be maintained by the LH power deposited around the foot point of ITB up to 25 sec, which corresponds to more than 100 times of current diffusion time, τL/R. ITB is terminated by the reduction of current drive efficiency caused by metal impurities accumulation. In some condition, self-organized slow sawtooth oscillations (SSSO) of plasma current, density, temperature, and so on with the period comparable to the current diffusion time have been also observed during ITB discharge. The oscillation has the capability of particle exhaust, as the result, it may play an role in the avoidance of the impurity accumulation and the dilution in the future steady state fusion plasma with ITB, as the edge-localized mode in H-mode.

記述言語：英語  

記述言語：英語  

Magnetic sensorless sensing experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The horizontal position is calculated from the vertical field coil current and voltage without using signals of magnetic probes placed nearby a plasma. The calculations are focused on the ripple frequency component of the power supply. There is no drift problem with the time integration of magnetic probe signals. The error of the derived plasma position is lower than 2&#37; of the plasma minor radius.

DOI： 10.1088/1009-0630/6/5/006

記述言語：英語  

Characteristics of ion acceleration in the internal magnetic reconnection events (IRE) have been studied by means of a neutral particle energy analyzer (NPA) in Tokyo Spherical Tokamak (TST-2). The major and minor radii are 0.38 m and 0.25m, respectively. The magnetic field strength is 0.3T and the maximum plasma current is up to 140 kA. The electron and ion temperatures are 0.4-0.5 keV and 0.1 keV, respectively and the electron density is ~1x1019 m-3. The NPA can be scanned toroidally from q = 74&deg; (cw) to q = 114&deg; (ccw), where q = 90&deg; corresponds to the perpendicular sightline. The direction of the plasma current is cw. The NPA signals are digitized at every 50 ms. The NPA is calibrated in the energy range of 0.1 keV < E < 8.4 keV. When the IRE occurs, it is observed that the plasma current increases by ~ 20&#37; and the loop voltage drops from 0.6 V to-5 V for ~ 0.1 ms. The enhanced charge exchange flux is observed by more than one order of magnitude at ~ 1 keV for this reconnection phase. The ion temperature increases by 80 eV at IREs. The angle q dependence of increment of Ti shows that DTi (q = 74&deg;) is higher than that for q = 114&deg;. This observation suggests that an ion is accelerated initially in the direction of magnetic field lines. The time evolution of the ion distribution function is simulated with a Fokker-Planck code taking into account the electric field effects.

記述言語：英語  

記述言語：英語  

記述言語：英語  

The longest tokamak discharge, with a duration of 11406 s (3 It 10 min), has been achieved. The global particle balance has been investigated. In the longest discharge, the global balance between the particle absorption and release of the wall was achieved at t similar to 30 min. After that, the plasma density was maintained by the recycling flux alone until the end of the discharge. The maximum wall inventory is about 3.6 x 10(20) H at t similar to 30 min, but it is finally released from the wall at the end of the discharge. The hydrogen release seems to be caused by the temperature increase in the whole toroidal area of the main chamber. Moreover, it has been observed that there is a large difference between the properties of wall recycling in the continuous gas feed case (i.e. static condition) and in the additional gas puff case (i.e. dynamic condition). In the static condition, the effective particle confinement time increases to similar to 10 s during the I min discharge and it increases to similar to 100 s before the global balance in the longest discharge. In the dynamic condition, the decay time of the electron density just after the gas puff, i.e. the effective particle confinement time, is constant at 0.2-0.3 s during the discharge. The large difference in the effective particle confinement time between the static and dynamic conditions seems to be caused by the reduction in the recycling coefficient due to the enhanced wall pumping resulting from the additional gas puff.

DOI： 10.1088/0029-5515/44/7/001

記述言語：英語  

記述言語：英語  

A remote steering antenna designed as a symmetric direction antenna is prepared for Electron Cyclotron Heating (ECH) and Current Drive (ECCD) experiments on the TRIAM-IM tokamak. The steering angles of 8-19 degrees are available in the design, in addition to the angle near 0 degrees. The antenna is tested at both of high and low power levels. The fraction of radiated power to injected power at the antenna is roughly evaluated at the low power test. The angle radiated from the antenna is checked from both of the intensity and phase measurements at the low power test. Several percentage losses at the antenna are evaluated at the high power test. The antenna system is evaluated to work correctly.

記述言語：英語  

Four types of plasma current ramp-up experiments in full non-inductively lower hybrid current driven (LHCD) plasmas were executed in TRIAM-1M: (1) current start-up by a combination of electron cyclotron resonance heating (ECRH) and LHCD, (2) tail heating by additional LHCD, (3) bulk heating by ECRH and (4) spontaneous ramp-up by a transition to enhanced current drive (ECD) mode. The time evolutions of plasma current during four types of ramp-up phase were adjusted by a simple model with two different time constants, which are a time defined by the total current diffusion time and a time constant for improving the current drive efficiency. In the case of (1) and (4), the latter time constant is significant during the current ramp-up phase. The improvement in the current drive efficiency in the ECD mode is likely to be caused by the increase in the effective refractive index along the magnetic field of the lower hybrid wave.

DOI： 10.1088/0029-5515/44/2/019

記述言語：英語  

記述言語：英語  

記述言語：英語  

A new operational scenario of advanced tokamak formation was demonstrated in the JT-60U tokamak. This was accomplished by electron cyclotron and lower hybrid waves, neutral beam injection, and the loop voltage supplied by the vertical field and shaping coils. The Ohmic heating (OH) solenoid was not used but a small inboard coil (part of the shaping coil), providing less than 20% of total poloidal flux, was used. The plasma thus obtained had both internal and edge transport barriers, with an energy confinement time of 1.6 times H-mode scaling, a poloidal beta of 3.6, and a normalized beta of 1.6, and a large bootstrap current fraction (>90%). This result opens up a possibility to reduce, and eventually eliminate, the OH solenoid from a tokamak reactor, which will greatly improve its economic competitiveness.

DOI： 10.1103/PhysRevLett.92.035001

記述言語：英語  

An integrated advanced scenario of tokamak formation without the use of an Ohmic heating (OH) solenoid in the JT-60U tokamak was investigated. The OH solenoid was not used but the electron cyclotron, lower hybrid waves, neutral beam injection and the loop voltage provided both the poloidal flux and vertical field. The plasma generated from this scenario had both internal and edge transport barriers with an energy confinement of 1.6 times H-mode scaling, poloidal beta of 3.6 and large bootstrap current fraction. The results show improved economic competitiveness of tokamak reactor without the use of OH solenoid.

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

An overview of steady state tokamak studies in TRIAM-IM (R-0 = 0.8 m, a x b = 0.12 m x 0.18 m and B = 8 T) is presented. The current ramp-up scenario without using centre solenoid coils is reinvestigated with respect to controllability of the current ramp-up rate at the medium density region of (1-2) x 10(19) m(-3). The plasma is initiated by ECH (fundamental o-mode at 170 GHz with 200 kW) at B = 6.7 T, and the ramp-up rate below the technical limit of 150 kA s(-1) for ITER can be achieved by keeping the LH power less than 100 kW during the current ramp-up phase. The physics understanding of the enhanced current drive (ECD) mode around the threshold power level has progressed from a viewpoint of transition probability. A transition frequency, f(trans), for the ECD transition is determined as a function Of P-CD. At similar to70 kW no transition occurs for an f(trans), value of similar to0.017 Hz, meaning almost zero transition probability. With increasing P-CD > P-th, f(trans) increases up to 10 Hz, and the transition tends to occur with high probability. The record value of the discharge duration is updated to 3 h 10 min in a low (n) over tilde (e) similar to 1 x 10(18) m(-3) and low power (<10 kW) discharge. The global particle balance in long duration discharges is investigated, and the temporal change in wall pumping rate is determined. Although the density was low, the gas supply had to be stopped at 30 min after the plasma initiation to maintain the density constant. After that the density was sustained by the recycling flux alone until the end of the discharges. In addition to the recycling problem, in the high power and high density experiments, the localized PWI affects the SSO of the tokamak plasma. The effects of enhanced influx of metal impurities (Fe, Cr, Ni, Mo) on sustainment of the high performance ECD plasma are investigated. In order to evaluate the helium bombarding effects on the plasma facing component and hydrogen recycling in the future burning plasma, microscopic damage of metals exposed to long duration helium discharges was studied. The total exposure time was 128 s. From thermal desorption experiments for the specimens the amount of retained helium was evaluated as 3.9 x 10(20) He m(-2) and the scale length to be similar to1 mm in the SOL.

DOI： 10.1088/0029-5515/43/12/006

記述言語：英語  

Sensorless sensing experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The horizontal position is calculated from the vertical field coil current and voltage in two ways. The calculations are made focusing on the low frequency component and the ripple frequency component of the power supply. In the latter case, there is no drift problem and the error is lower than 2&#37; of the plasma minor radius. (C) 2003 Elsevier Science B.V. All rights reserved.

DOI： 10.1016/S0920-3796(03)00308-9

記述言語：英語  

記述言語：英語  

A fully superconducting tokamak named JT-60SC is designed for the modification programme of JT-60 to enhance economical and environmental attractiveness in tokamak fusion reactors. JT-60SC aims at realizing high-beta steady-state operation in the use of low radio-activation ferritic steel in a low v* and rho* regime relevant to the reactor plasmas. Objectives, research issues, plasma control schemes and a conceptual design for JT-60SC are presented.

DOI： 10.1088/0029-5515/43/7/314

記述言語：英語  

The operation of tokamak devices is based on the principle of the transformer, so it has been considered that pulse operation of such devices is inevitable. But the superconducting high-field tokamak TRIAM-1M has been demonstrated continuous operation for last 15 years and succeeded in sustaining 3 hours of constant operation in 2001. Fusion reactors of the future will use the center solenoid (CS) coil only in the ramp-up phase of the plasma current. If the coil is removed from the tokamak device, the device can be simplified and its cost will be decreased. In TRIAM-1M experiments, a perfect CS-free ramp-up was demonstrated and succeeded in sustaining the plasma current longer than 30 s. In this paper, the operation scenario, especially plasma position control, and the results of the analysis of plasma control are described.

記述言語：英語  

A large difference between properties of wall recycling in the continuous gas feed case (i.e. static condition) and the additional gas puff case (i.e. dynamic condition) has been observed. In the static condition, the effective particle confinement time, tau(p)(*), increases almost linearly to about 10 s during the I min discharge. In the dynamic condition, tau(p)(*) is 0.2-0.3 s during the I min discharge. This difference of tau(p)(*) is also confirmed in the ultra-long discharge. tau(p)(*) in the static condition becomes similar to100 s before the global balance between particle absorption and release of the wall is achieved at t similar to30 min. tau(p)(*) in the dynamic condition is, however, still on the order of similar to0.3 s. The large difference between tau(p)(*) in the static and dynamic conditions is attributed to a reduction in the recycling coefficient due to enhanced wall pumping resulting from the gas puff. (C) 2003 Elsevier Science B.V. All rights reserved.

DOI： 10.1016/S0022-3115(02)01390-9

記述言語：英語  

An analytical model of fluctuation reflectometry is developed, which is capable of handling the plasma profiles of arbitrary shape and curvature. The experimental profiles are piece-wise approximated by the functions, which allow for the integration of the full-wave O-mode equation. The model is applied to the reflectometry of the JT-60U tokamak plasma to provide a preliminary estimation of the fluctuation amplitude and spectral width. (C) 2003 American Institute of Physics.

DOI： 10.1063/1.1538355

記述言語：英語  

An analytical model of fluctuationreflectometry is developed, which is capable of handling the plasma profiles of arbitrary shape and curvature. The experimental profiles are piece-wise approximated by the functions, which allow for the integration of the full-wave O-mode equation. The model is applied to the reflectometry of the JT-60U tokamak plasma to provide a preliminary estimation of the fluctuation amplitude and spectral width.

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

記述言語：英語  

Research activities of the Japanese tokamaks JT-60U, JFT-2M, and TRIAM-1M are described. The recent JT-60 program is focused on the establishment of a scientific basis of advanced steady-state operation. Plasma performance in transient and quasi steady states has been significantly improved, utilizing reversed shear and weak shear (high-p) ELMy H-modes characterized by both internal and edge transport barriers and high bootstrap current fractions. Development of each key issue for advanced steady-state operation has also been advanced. Advanced and basic research of JFT-2M has been performed to develop high-performance tokamak plasma as well as the structural material for a fusion reactor. Toroidal field ripple reduction with ferritic steel plates outside the vacuum vessel is successfully demonstrated. No adverse effects to the plasma were observed with poloidal fields inside the vacuum vessel (partial covering). Preparation is in progress for full-scale testing of the compatibility of the ferritic steel wall (full covering) with plasma. A heavy ion beam probe has been installed to study H-mode plasmas. Compact toroid (CT) injection experiments are performed, showing deep CT penetration into the core region of the H-mode. The TRIAM project has investigated steady-state operation and high-performance plasma of a tokamak with the high toroidal magnetic field superconducting tokamak. Four important contributions in the fields of fusion technology of superconducting tokamaks, steady-state operation, high-performance plasma, and startup of plasma current without the assistance of center solenoid coils have been achieved on TRIAM-1M, especially regarding steady-state operation by realization of a discharge for >3 h.

記述言語：英語  

Controllability of current profiles in long duration discharges is studied by two kinds of combination of wave spectra. First, by superposition of a higher N-parallel to spectrum wave at 2.45 GHz on a target plasma driven by a lower N-parallel to one at 8.2 GHz, a hollow j (r) profile has been achieved and sustained for 20 s. The current profile can be well controlled below a threshold power of 14 kW by varying the power at 2.45 GHz. A spontaneous transition phenomenon to a peaked j (r) profile, however, has occurred above the threshold power even after a reasonable steady state condition had been obtained. Second, using two oppositely travelling waves at 8.2 GHz the total current is clearly reduced, and the j (r) profile becomes peaked when the backward travelling lower hybrid waves (BWs) are superposed on the forward lower hybrid wave at a low power of 20 kW. When the BW power is increased above 27 kW, further current reduction is suppressed, and above 34 kW the direction of the current driven by the BWs is completely reversed and the j (r) profile becomes broad. Thus, current modification by BWs shows a highly non-linear behaviour with respect to the BW power.

DOI： 10.1088/0029-5515/42/3/315

記述言語：英語  

記述言語：英語  

Recycling and wall pumping have been studied comparing low (approximate to10(18) m(-3)) and high (approximate to10(19) m(-3)) density long duration plasmas in TRIAM-1M. The recycling coefficient of each plasma increases with time. There exist two time constants in the temporal evolution of the recycling coefficient. One is a few seconds and the other is about 30 s. The wall pumping rates of low and high density plasmas are evaluated to be approximate to1.5 x 10(16) atoms m(-2) s(-1) and approximate to4 x 10(17) atoms m(-2) s(-1), respectively. The difference may be caused by the total number of diffused ions and the charge exchange neutral flux with an energy of less than 0.7 keV. In an ultra-long discharge (approximate to70 min), the recycling coefficient becomes 1 or more before again decreasing below 1, i.e. the wall repeats the process of being saturated and refreshed. In high power and high density experiments, wall saturation phenomena have been observed. The discharge duration limited by wall saturation decreases with an increase in density.

DOI： 10.1088/0029-5515/42/2/307

記述言語：英語  

記述言語：英語  

Research activities of the Japanese tokamaks JT-60U, JFT-2M, and TRIAM-1M are described. The recent JT-60 program is focused on the establishment of a scientific basis of advanced steady-state operation. Plasma performance in transient and quasi steady states has been significantly improved, utilizing reversed shear and weak shear (high-β_p) ELMy H-modes characterized by both internal and edge transport barriers and high bootstrap current fractions. Development of each key issue for advanced steady-state operation has also been advanced. Advanced and basic research of JFT-2M has been performed to develop high-performance tokamak plasma as well as the structural material for a fusion reactor. Toroidal field ripple reduction with ferritic steel plates outside the vacuum vessel is successfully demonstrated. No adverse effects to the plasma were observed with poloidal fields inside the vacuum vessel (partial covering). Preparation is in progress for full-scale testing of the compatibility of the ferritic steel wall (full covering) with plasma. A heavy ion beam probe has been installed to study H-mode plasmas. Compact toroid (CT) injection experiments are performed, showing deep CT penetration into the core region of the H-mode. The TRIAM project has investigated steady-state operation and high-performance plasma of a tokamak with the high toroidal magnetic field superconducting tokamak. Four important contributions in the fields of fusion technology of superconducting tokamaks, steady-state operation, high-performance plasma, and startup of plasma current without the assistance of center solenoid coils have been achieved on TRIAM-1M, especially regarding steady-state operation by realization of a discharge for > 3 h.

DOI： 10.13182/FST02-A210

記述言語：英語  

記述言語：英語  

The enhanced current drive (ECD) efficiency mode, which is characterized by a spontaneous increase of current drive efficiency eta (CD) from (0.3-0.4) x 10(19) A/W m(-2) to (0.6-1.0) x 10(19) A/W M(-2), is observed in long pure LHCD plasmas on TRIAM-1M. The energy confinement time is also improved due to the increase of line averaged electron density, and of the ion and electron temperatures. The current drive efficiency is proportional to the electron density. The transition to ECD mode occurs at a critical density, which depends slightly on the refractive index in the toroidal direction N(parallel to) of the injected wave.

DOI： 10.1088/0029-5515/41/11/302

記述言語：英語  

The main purpose of TRIAM-1M (R-0 = 0.8 m, a x b = 0.12 m x 0.18 m, B = 8 T) is to study the route towards a high field compact steady state fusion reactor. In the advanced steady state operation programme, a heating mechanism for the high ion temperature mode with an internal transport barrier has been studied, an enhanced current drive mode in an extended (higher power and higher density) operation regime has been obtained, current density profile control experiments using multicurrent drive systems have been performed and the effects of wall recycling, wall pumping and wall saturation on particle control have been investigated.

DOI： 10.1088/0029-5515/41/10/315

記述言語：英語  

記述言語：英語  

Wall conditioning using electron cyclotron resonance-discharge cleaning (ECR-DC) was carried out for 35 h in a hydrogen plasma on superconducting tokamak, TRIAM-1M after ventilation of vacuum vessel for 2 months. The baking-out on extension ports was carried out before ECR-DC. As the wall of plasma chamber was cooled by water during ECR-DC about 25 degreesC, the pure effect of ECR-DC could be investigated. The ECR plasma of the electron density similar to 2x10(10) cm(-3) and electron temperature similar to6 eV was maintained by the microwave power up to 1 kW provided by a magnetron of frequency in 2.45 GHz. The decay time of the partial pressure of mass number 18 (H2O) and 28 (C2H4 and/or CO) were monitored during ECR-DC. The time evolution shows that oxygen and carbon was well removed from the surface of the wall by ECR-DC. The plasma resistivity of the first ohmic heated (OH) plasma just after the ECR-DC was the same as that of the final OH plasma after exposing the wall to tokamak plasmas for more than 7000 s. Intensity of the O II line, which was investigated in lower hybrid current drive (LHCD) plasma, did not change during an experimental campaign. These indicate that the wall conditioned by only ECR-DC is good in both OH and LHCD plasmas. A quantitative model of the efficiency for ECR-DC is presented. The model can explain the time evolution of partial pressure. From this model the duration required to clean up using ECR-DC was derived. This result may help in the design of the pump in large machines. (C) 2001 Elsevier Science B.V. All rights reserved.

DOI： 10.1016/S0920-3796(00)00434-8