Language：Japanese   Publishing type：Research paper (scientific journal)  

QUEST is a middle-sized spherical tokamak, which has a capability of long term plasma discharge. Thus, the QUEST is suitable for the study of steady state operation such as energy and particle balances and particle recycling. For the sustainment of steady-state plasma, it may be required such as high performance control systems, various sensors and actuators, technics for real-time calculation methods, and so on. In this paper, the present status of control systems on QUEST are presented, and its perspectives are also described. The central control system will be improved with easy comprehensible user interface and Ethernet environment. And, the plasma control system identify the plasma position and its current with hall sensors in real-time continuously without drift errors. The plasma on QUEST will be controlled integrally with the cooperation of these systems.

Language：English   Publishing type：Research paper (scientific journal)  

The plasma control system (PCS) of QUEST is developed according to the progress of QUEST project. Sinceone of the critical goals of the project is to achieve the steady-state operation with high temperaturevacuum 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 ofreflective memories (RFMs). The PCS controls the plasma edge position with the real-time identificationof 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 H signal filtered by a digital low-pass filterare proposed and suitable for the steady-state operation on QUEST. The present status and prospect ofthe PCS are presented with recent topics.

DOI： http://dx.doi.org/10.1016/j.fusengdes.2016.04.016

Repository Public URL： https://hdl.handle.net/2324/7183072

Language：Japanese   Publishing type：Research paper (scientific journal)  

プラズマを高温・高密度で真空容器内に閉じ込めるトカマク装置において、真空容器外に設置したホール素子を用いて、真空容器内のプラズマ電流およびプラズマ位置を評価する方法の提案をおこない、その結果を球状トカマク装置QUESTに適用して、他の磁気計測による方法と比較して適切に評価が行えることを確認した。

Language：English   Publishing type：Research paper (scientific journal)  

The plasma control system (PCS) of QUEST was a centralized system, which lost its scalability because ofthe overload imposed on its central processing unit (CPU) of the PCS, making it impossible to add newfunctions. Thus, the PCS is distributed into a main workstation (WS) and subsystem (SS) with a reflectivememory (RFM) in order to share data between these systems so as to mitigate the load on each system.As a result, 128 double-precision floating-point numbers (DBLs) can be transferred from the SS to the WSwith a maximum latency of 250 s. The WS and the SS each have quad-core CPUs, and tasks are executedin parallel. Although one of the four cores is intermittently occupied by up to 90% by this transaction, theoccupation is normally 60%. A time correction procedure is used to map the recorded data sets on theWS and the SS to a common time base by referring to the time difference between two systems.

Repository Public URL： https://hdl.handle.net/2324/7183071

Language：English   Publishing type：Research paper (scientific journal)  

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

Repository Public URL： https://hdl.handle.net/2324/7183070

Language：English   Publishing type：Research paper (scientific journal)  

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

Repository Public URL： https://hdl.handle.net/2324/7183069

Language：English   Publishing type：Research paper (scientific journal)  

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

Repository Public URL： https://hdl.handle.net/2324/7183068

Language：English   Publishing type：Research paper (scientific journal)  

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

Repository Public URL： https://hdl.handle.net/2324/7183065

Language：English   Publishing type：Research paper (scientific journal)  

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

Repository Public URL： https://hdl.handle.net/2324/7183066

Language：English   Publishing type：Research paper (scientific journal)  

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

Repository Public URL： https://hdl.handle.net/2324/7183067

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：Japanese   Publishing type：Research paper (bulletin of university, research institution)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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% of the plasma minor radius.

DOI： 10.1088/1009-0630/6/5/006

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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% of the plasma minor radius. (C) 2003 Elsevier Science B.V. All rights reserved.

DOI： 10.1016/S0920-3796(03)00308-9

DOI： 10.1088/1361-6587/ad9521

Web of Science

Publisher：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

Publisher：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

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Scopus

Language：English   Publisher：Transdisciplinary Research and Education Center for Green Technologies, Kyushu University  

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

Publisher：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

Publisher：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

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Scopus

Publisher：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

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Scopus

Publisher：Nuclear Fusion  

Spatially resolved rotational temperature of ground state hydrogen molecules desorbed from plasma-facing surface was measured in QUEST, LTX-β, and DIII-D tokamaks, and the increases of the rotational temperature with the surface temperature and due to collisional-radiative processes in the plasmas were evaluated. The increase due to collisional-radiative processes was calculated by solving rate equations considering electron and proton collisional excitation and deexcitation and spontaneous emission. The calculation results suggest a high sensitivity for the rotational temperature to electron and proton densities, but a negligible sensitivity to the electron, proton, and surface temperatures. In the three tokamaks with different plasma parameters and plasma-facing surface materials, the spatial profile of the rotational temperature was estimated using Fulcher-α emission lines (600-608 nm). In QUEST, the spatial profile of the rotational temperature was estimated from spatially resolved spectra. In the other tokamaks, the rotational temperature was evaluated assuming a single point emission with a location determined from the Fulcher-α emission profile as measured with a filtered camera. In metal-walled devices QUEST and LTX-β, the rotational temperature increased with the surface temperature, and the calculated collisional-radiative increase is consistent with measured increase assuming that the rotational temperature at the surface is approximately 500-600 K higher than the surface temperature. In DIII-D with carbon walls, a larger collisional-radiative increase than the other tokamaks was observed because of the higher density leading to a large difference from the calculated increase compared to the other smaller tokamaks. Measurement of the Fulcher-α emission profile with higher spatial resolution in DIII-D may reduce the difference and reveal the effect of the surface temperature on the rotational temperature. These results show the increases in the rotational temperature with the surface temperature and due to the collisional-radiative processes.

DOI： 10.1088/1741-4326/acd4d1

Web of Science

Scopus

Language：English   Publisher：IOP Publishing  

Spatially resolved rotational temperature of ground state hydrogen molecules desorbed from plasma-facing surface was measured in QUEST, LTX-β, and DIII-D tokamaks, and the increases of the rotational temperature with the surface temperature and due to collisional-radiative processes in the plasmas were evaluated. The increase due to collisional-radiative processes was calculated by solving rate equations considering electron and proton collisional excitation and deexcitation and spontaneous emission. The calculation results suggest a high sensitivity for the rotational temperature to electron and proton densities, but a negligible sensitivity to the electron, proton, and surface temperatures. In the three tokamaks with different plasma parameters and plasma-facing surface materials, the spatial profile of the rotational temperature was estimated using Fulcher-α emission lines (600–608 nm). In QUEST, the spatial profile of the rotational temperature was estimated from spatially resolved spectra. In the other tokamaks, the rotational temperature was evaluated assuming a single point emission with a location determined from the Fulcher-α emission profile as measured with a filtered camera. In metal-walled devices QUEST and LTX-β, the rotational temperature increased with the surface temperature, and the calculated collisional-radiative increase is consistent with measured increase assuming that the rotational temperature at the surface is approximately 500–600 K higher than the surface temperature. In DIII-D with carbon walls, a larger collisional-radiative increase than the other tokamaks was observed because of the higher density leading to a large difference from the calculated increase compared to the other smaller tokamaks. Measurement of the Fulcher-α emission profile with higher spatial resolution in DIII-D may reduce the difference and reveal the effect of the surface temperature on the rotational temperature. These results show the increases in the rotational temperature with the surface temperature and due to the collisional-radiative processes.

CiNii Research

Language：English   Publisher：The Japan Society of Plasma Science and Nuclear Fusion Research  

<p>Non-inductive plasma current start-up experiments have been conducted using 2^ndharmonic electron-cyclotron resonance heating (ECRH) on the QUEST spherical tokamak. Sudden or abrupt changes in plasma current, referred to as sudden change events (SCEs), have been frequently observed during these experiments. To investigate these events, fast magnetic activity was measured globally by installing arrays of magnetic pick-up coils inside the vacuum vessel on QUEST. Analysis revealed that all observed SCEs could be classified into three categories based on their magnetic characteristics. In one type of event, called SCE–III, the plasma current suddenly decreases by up to 50% instantaneously, occurring in a wide discharge region achieved using 2^ndECRH start-up on QUEST. These SCE–III events may potentially lead to disruptions in future discharges at higher powers.</p>

DOI： 10.1585/pfr.18.2402066

Web of Science

Scopus

CiNii Research

Language：English   Publishing type：Research paper (scientific journal)  

Various scenarios of electron cyclotron heating can be explored in a spherical tokamak with a high toroidal field. On the QUEST spherical tokamak, doubling the toroidal field results in the emergence of the fundamental resonance layer for 28 GHz electron cyclotron waves. The second-harmonic resonance layer shifts toward the plasma central area. To increase the coil current from 50 to 100 kA, a current source is designed with 125 modules in parallel. The module power supply comprises a bank of lithium-ion capacitors and a DC–DC buck converter. In the bench test of a single module, the current was found to increase linearly and maintained the flat-top as programmed.

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

Publisher：Fusion Engineering and Design  

Various scenarios of electron cyclotron heating can be explored in a spherical tokamak with a high toroidal field. On the QUEST spherical tokamak, doubling the toroidal field results in the emergence of the fundamental resonance layer for 28 GHz electron cyclotron waves. The second-harmonic resonance layer shifts toward the plasma central area. To increase the coil current from 50 to 100 kA, a current source is designed with 125 modules in parallel. The module power supply comprises a bank of lithium-ion capacitors and a DC–DC buck converter. In the bench test of a single module, the current was found to increase linearly and maintained the flat-top as programmed.

DOI： 10.1016/j.fusengdes.2023.113794

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Scopus

Publisher：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

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Scopus

Language：English   Publisher：The Japan Society of Plasma Science and Nuclear Fusion Research  

<p>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.</p>

DOI： 10.1585/pfr.18.2405035

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Scopus

CiNii Research

Publisher：Plasma Science and Technology  

A power-supply system was developed for Ohmic heating (OH) to double × 1018 the amount of change magnetic flux in the primary central solenoid (CS) on the QUEST spherical tokamak. Two power supplies are connected with stacks of insulated-gate bipolar transistors, and sequentially operated to generate positive and negative CS currents. This bipolar power-supply system is controlled via a field-programmable gate array, which guarantees the safety of the entire system operation. The new OH system, assisted by electron cyclotron heating, enables the stable generation of plasma currents exceeding 100 kA. Moreover, the achieved electron density over the wide range in the major radial direction exceeds the cut-off density for one of the high-power microwave sources in QUEST. This strategy yields target plasmas for future experiments with the electron Bernstein wave.

DOI： 10.1088/2058-6272/acafc2

Web of Science

Scopus

Language：English   Publisher：The Japan Society of Plasma Science and Nuclear Fusion Research  

<p>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.</p>

DOI： 10.1585/pfr.18.1405012

Web of Science

Scopus

CiNii Research

Language：English   Publishing type：Research paper (scientific journal)   Publisher：Journal of Fusion Energy  

Transient coaxial helicity injection (CHI) is a promising candidate for solenoid-free plasma current start-up in a low-aspect-ratio tokamak in support of developing fully non-inductive scenarios. The aim of the transient CHI research on QUEST is to develop a reactor-relevant CHI design. On QUEST, a CHI discharge is initiated by driving current along magnetic field lines that connect an electrically insulated electrode plate (which is referred to as a bias electrode) to a vessel component at the bottom region of the spherical tokamak. In the first application of CHI on QUEST, the electrically insulated electrode plate was biased with respect to the outer vessel in a configuration referred to as low-field-side (LFS) injection. To maintain a narrow injector flux footprint width throughout the discharge, QUEST is now developing a high-field-side (HFS) injection configuration, in which the electrically insulated electrode plate is biased with respect to the inner vessel components. Through the implementation of a CHI-dedicated gas injection system, studies in the HFS injection configuration have now demonstrated good magnetic flux evolution into the vacuum vessel. Toroidal currents up to 43 kA have been achieved, and the generated current has increased in proportion to the magnetic flux connecting the electrodes. These results which show agreement with the CHI-scaling suggest that much higher levels of toroidal current can be generated on QUEST in an optimized CHI system in which the magnetic flux connecting the CHI electrodes is further increased.

DOI： 10.1007/s10894-022-00338-4

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Scopus

Language：English   Publishing type：Research paper (scientific journal)  

A heavy ion beam probe (HIBP) has been designed for the QUEST spherical tokamak to measure plasma turbulence and the profiles of electric potential profiles. Using a cesium ion beam with an energy of several 10 keV, the observable region covers most of the upper half of the plasma. Although the probe beam is deflected by the poloidal magnetic field produced by plasma current and poloidal coil currents, it can be detected under plasma current up to 150 kA by modifying the trajectories with two electrostatic sweepers. According to the numerical estimation of the intensity of the detected beam, sufficient signal intensity for measuring plasma turbulence can be obtained over almost the measurable area when the electron density is up to 1 × 1019 m−3, which is larger than the cut-off density of electron cyclotron heating in QUEST. The performance of the designed HIBP is sufficient to explore the mechanisms of heat and particle transport in magnetically confined plasmas, including the influence of plasma wall interactions, which is a goal of the QUEST project.

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

Publisher：IEEE Transactions on Plasma Science  

Transient coaxial helicity injection (T-CHI), a method first developed on the small helicity injected torus (HIT-II) experiment and then validated on the much larger National Spherical Torus Experiment (NSTX) device, is a method to initiate an inductive-like tokamak plasma discharge without reliance on the central solenoid. A CHI discharge is initiated by driving current along magnetic flux that connects the inner and outer divertor plates on one end of the tokamak. To permit this, on both HIT-II and NSTX, toroidal ceramic insulators were used to electrically separate the inner and outer vessel components. The use of such large toroidal vacuum insulators may not be easy to implement in reactors. To address this issue, the Q-shu University experiments with steady-state spherical tokamak (ST) (QUEST) is developing a reactor-relevant CHI configuration in which one of the divertor plates is electrically insulated from the rest of the vessel. The first application of T-CHI on QUEST biased the CHI electrode to the outer vessel. While the CHI discharges could be easily generated, it was found that as the discharge filled the vessel, the separation distance between the injector magnetic flux footprints widened, a condition that is not favorable for the generation of closed flux surfaces. Biasing the electrode to the inner wall is a configuration similar to that used on NSTX and HIT-II, but initial testing in this configuration has proved to be challenging. The design described here overcomes the present limitation by locating the CHI electrode much closer to the CHI injector flux coil and using an NSTX-like gas injection manifold to enable high-field-side T-CHI startup on QUEST. The concepts described in this article should also benefit the future implementation of T-CHI systems in other tokamaks and spherical tokamaks.

DOI： 10.1109/TPS.2022.3193069

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Scopus

Language：English   Publishing type：Research paper (scientific journal)  

Transient coaxial helicity injection (T-CHI), a method first developed on the small helicity injected torus (HIT-II) experiment and then validated on the much larger National Spherical Torus Experiment (NSTX) device, is a method to initiate an inductive-like tokamak plasma discharge without reliance on the central solenoid. A CHI discharge is initiated by driving current along magnetic flux that connects the inner and outer divertor plates on one end of the tokamak. To permit this, on both HIT-II and NSTX, toroidal ceramic insulators were used to electrically separate the inner and outer vessel components. The use of such large toroidal vacuum insulators may not be easy to implement in reactors. To address this issue, the Q-shu University experiments with steady-state spherical tokamak (ST) (QUEST) is developing a reactor-relevant CHI configuration in which one of the divertor plates is electrically insulated from the rest of the vessel. The first application of T-CHI on QUEST biased the CHI electrode to the outer vessel. While the CHI discharges could be easily generated, it was found that as the discharge filled the vessel, the separation distance between the injector magnetic flux footprints widened, a condition that is not favorable for the generation of closed flux surfaces. Biasing the electrode to the inner wall is a configuration similar to that used on NSTX and HIT-II, but initial testing in this configuration has proved to be challenging. The design described here overcomes the present limitation by locating the CHI electrode much closer to the CHI injector flux coil and using an NSTX-like gas injection manifold to enable high-field-side T-CHI startup on QUEST. The concepts described in this article should also benefit the future implementation of T-CHI systems in other tokamaks and spherical tokamaks.

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

Language：English   Publisher：Elsevier B.V.  

In order to generate high-performance plasma for future fusion power generation, it is desirable to keep high quality vacuum during experiment. Mass spectrometer is commonly used to monitor the vacuum quality and to record the amount of atoms and molecules in the vacuum vessel. Leak is the most serious accident to avoid that can nullify an experiment and even harm researchers. Detecting leaks are ever more important since it can be easily overlooked, e.g., when the deterioration in the vacuum degree is modest. This forces the researcher to carefully observe the vacuum and mass spectrometer data. This article presents a way to suggest potential leaks in the vacuum vessel by analyzing mass spectrometer data. This is done by utilizing the Euclidean distance between composition ratios at different times for the clustering using the daily composition ratio. We show that our cluster analysis is an effective way of separating these two cases, which results in a semi-automatic determination of leaks is more efficient than the current norm, which is to check many measures to find a small abnormality in the data manually. We plan further model improvements for long-term evaluation.

DOI： 10.1016/j.fusengdes.2022.113199

Web of Science

Scopus

CiNii Research

Language：English   Publisher：The Japan Society of Plasma Science and Nuclear Fusion Research  

<p>The space vector of an output voltage can be rotated smoothly in a matrix converter. However, a zero-sequence component appears causing problems in the load, such as a motor. A quaternion is a four-dimensional hypercomplex number with an imaginary part that can simultaneously represent and deal with three-phase voltages. In addition, the quaternion is expressed in the exponential form; thus, it can easily represent the space vector rotation. Two zero configurations were used to optimize the ripple characteristics in fictitious pulse-width-modulated voltage-source inverter. The two zero configurations are used to remove the zero-sequence component from the matrix converter. The quaternion can be differentiated in time as well as rotate in space. Therefore, it is used to analyze transient phenomena in the matrix converter's rise-up and rise-down phases, and the switching's transition phase.</p>

DOI： 10.1585/pfr.17.2405025

Web of Science

Scopus

CiNii Research

Language：English   Publishing type：Research paper (scientific journal)  

The space vector of an output voltage can be rotated smoothly in a matrix converter. However, a zero-sequence component appears causing problems in the load, such as a motor. A quaternion is a four-dimensional hypercomplex number with an imaginary part that can simultaneously represent and deal with three-phase voltages. In addition, the quaternion is expressed in the exponential form; thus, it can easily represent the space vector rotation. Two zero configurations were used to optimize the ripple characteristics in fictitious pulse-width-modulated voltage-source inverter. The two zero configurations are used to remove the zero-sequence component from the matrix converter. The quaternion can be differentiated in time as well as rotate in space. Therefore, it is used to analyze transient phenomena in the matrix converter's rise-up and rise-down phases, and the switching's transition phase.

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

Language：English   Publishing type：Research paper (scientific journal)  

Noninductive plasma current start-up using 2nd harmonic electron cyclotron resonance heating (ECRH) with oblique radio frequency (RF) injection is demonstrated in a Q-shu University experiment with steady-state spherical tokamak. A strong transition was observed in the heating and plasma current ramp-up. The initial bulk electron heating regime exhibits Tebulk ∼ 140 eV and no hard x-ray (HXR) emission with a low Ip of ∼15 kA; it abruptly transitions to a regime that exhibits a low Tebulk of ∼10 eV and a strong HXR emission with a high Ip of ∼50 kA. This behavior is distinctly different from that observed in previous fundamental ECRH experiments. The mechanism of the heating and current drive transition are investigated considering wave power absorption and plasma power balance. The results indicate that the transition is caused by the favorable heating of tail electrons where the RF power absorption at the 2nd harmonic increases nearly linearly with Tetail, while the power transfer from the tail electrons to the bulk electrons decreases with 1/Tetail 0.5. This causes a rapid transition to a state with high Tetail while reducing Tebulk towards colder ion temperature. The understanding of the transition mechanism
helps to consider plasma current start-up using 2nd harmonic ECRH for tokamak reactors such as JT-60 SA and ITER.

DOI： https://doi.org/10.1088/1361-6587/ac1838

Language：English   Publishing type：Research paper (scientific journal)  

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 recy-cling 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 tem-perature, 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： https://doi.org/10.1016/j.nme.2021.101013

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

The parametric decay wave (PDW) caused by three-wave parametric decay process was measured in a plasma-injecting X-mode electron cyclotron wave (ECW) from the high field side (HFS) of the Q-shu University Experimental Steady-State Spherical Tokamak (QUEST). The intensity of the low-frequency PDW on the HFS X-mode injection was significantly enhanced over the O-mode ECW injection from the low field side (LFS), where the mode conversion to electron Bernstein wave (EBW) was not expected. As the comparison was executed using plasmas with the same magnetic field and injection power, the wave injection method was considered as the primary cause of the difference in the PDW excitation. The frequency range of the low-frequency PDW was consistent with that of the lower hybrid wave (LHW) range, which was expected to be excited during the mode conversion to EBW. The low-frequency PDW intensity evolved in response to the plasma density, plasma current and injection power. This observation suggests that the low-frequency PDW intensity is a reliable indicator for the efficient mode conversion to EBW.

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

Other Link： https://www.jstage.jst.go.jp/article/pfr/15/0/15_2402063/_pdf/-char/en

Language：English   Publishing type：Research paper (scientific journal)  

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

Other Link： http://www.jspf.or.jp/PFR/PDF2020/pfr2020_15-2402013.pdf

Language：English   Publishing type：Research paper (scientific journal)  

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% 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%, compared to 40% 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 BPF = 7.6 mT, plasma current (IP) 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 IP 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

Other Link： https://iopscience.iop.org/article/10.1088/1361-6587/ab6903

Language：English   Publishing type：Research paper (scientific journal)  

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

Other Link： https://iopscience.iop.org/article/10.1088/1741-4326/ab4c12

Language：English   Publishing type：Research paper (scientific journal)  

A two-dimensional (2D) soft X-ray (SXR) imaging system for obtaining tangential views of plasmaproduced SXR has been installed on the QUEST tokamak. The system comprises a chevron micro-channel
plate (MCP) and a phosphor screen, which together serve as an SXR detector and image intensifier, along with a
high-speed video camera for capturing and storing image data. The system can be used in imaging mode to obtain
high spatial- and time-resolution images or in photon counting mode at an ultra-high framing rate of 100 kHz to
obtain SXR energy spectra. Here, we propose a method of in situ energy calibration based on the transmission
characteristics of a 25-µm Be filter. The proposed method was used to obtain SXR spectra in the energy range
1.5 - 3.5 keV with a core electron temperature in agreement with electron temperatures measured by the Thomson
scattering method. Furthermore, oscillations based on the modification of plasma parameters and rapid increases
in impurity radiation were detected in imaging mode. This dual-mode-compatible method for obtaining 2D SXR
diagnostics is useful for the study of high temperature plasmas.

DOI： 10.1585/pfr.14.1402128

Other Link： http://www.jspf.or.jp/PFR/PDF2019/pfr2019_14-1402128

Language：English   Publishing type：Research paper (scientific journal)  

The fuel particle balance during long duration discharges in the Q-shu University Experiment with steady state spherical tokamak (QUEST) was investigated. QUEST has all-metal plasma facing walls (PFWs) that were temperature controlled during the experiments. The presence of a transport barrier for hydrogen (H) at the interface between a plasma-induced deposition layer and metallic substrate was confirmed by nuclear reaction analysis with exposing deuterium plasma. An effective method to evaluate global hydrogen flux to PFWs is proposed, taking advantage of the nature of wall saturation. The outgoing flux of fuel particles from the PFWs just after the plasma termination was proportional to the square of wall-stored H, which indicates that enhanced recombination of solved hydrogen played an essential role in dynamic retention and was in agreement with predictions from the H-barrier model. A simple calculation based on the combination of wall modelling and rate equations of the H states denoted a significant impact of wall modelling on the time response of the plasma density. Hence, a proper wall model including the effects of the deposition layer creating the H barrier is required to be developed, even for all-metal PFW devices.

DOI： 10.1088/1741-4326/ab1858

Other Link： https://iopscience.iop.org/article/10.1088/1741-4326/ab1858

Language：English   Publishing type：Research paper (scientific journal)  

This research investigated fuel particle balance during long duration discharge in an all-metal plasma facing wall (PFW) through intensive QUEST execution. A simple wall model including the plasma-induced deposition layer that creates hydrogen (H) barriers, called the H barrier model, was established. A simple calculation, based on a combination of H state rate equations and the H barrier model, was applied to real plasma in the early phase of its longest discharge. The model accurately reconstructed the evolutions of electron density and wall-stored H over time, proper values are chosen for the parameters that are difficult to determine experimentally. Comparative calculations that used the H barrier and a fully reflective models, predicted significant impacts of wall models on the plasma density time response and value of electron density, indicating that a proper wall model should be developed for all-metal PFW devices.

DOI： https://doi.org/10.1016/j.nme.2019.03.015

Other Link： https://www.sciencedirect.com/science/article/pii/S2352179118302242?via%3Dihub

Language：English   Publishing type：Research paper (scientific journal)  

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： https://doi.org/10.1016/j.fusengdes.2019.04.043

Other Link： https://www.sciencedirect.com/science/article/pii/S0920379619305770?via%3Dihub

Language：English   Publishing type：Research paper (scientific journal)  

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： https://doi.org/10.1016/j.fusengdes.2019.04.059

Other Link： https://reader.elsevier.com/reader/sd/pii/S0920379619305939?token=0FBA06FE59FB0F0A4D25E3A18C1DA47BFEA925269241FCBDB5AFB49806F68BBA0453DAF28E90EFA16ED06A30FBE9BB08&originRegion=us-east-1&originCreation=20210527052001

Language：English   Publishing type：Research paper (scientific journal)  

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： https://doi.org/10.1016/j.fusengdes.2019.02.027

Other Link： https://www.sciencedirect.com/science/article/pii/S0920379619302005?via%3Dihub

Language：English   Publishing type：Research paper (scientific journal)  

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 mid-plane 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 Hα 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 35 A in the LFS injection. The electron density ne predicted from the position of the upper hybrid resonance agreed with the line-averaged ne 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： https://doi.org/10.1585/pfr.14.1205038

Other Link： https://www.jstage.jst.go.jp/article/pfr/14/0/14_1205038/_pdf/-char/en

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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/B v )·(∂B v /∂R)
n=−(R/Bv)·(∂Bv/∂R)
, where R is the major radius and Bv 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 Te for RF-induced breakdown to realize an avalanche of electrons in the tokamak magnetic structure.

DOI： 10.1063/1.4985142

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

Fully non-inductive plasma start-up was successfully achieved by using a well-controlled microwave source on the spherical tokamak, QUEST. Non-inductive plasmas were maintained for approximately 3–5 min, during which time power balance estimates could be achieved by monitoring wall and cooling-water temperatures. Approximately 70%–90% of the injected power could be accounted for by calorimetric measurements and approximately half of the injected power was found to be deposited on the vessel wall, which is slightly dependent on the magnetic configuration. The power distribution to water-cooled limiters, which are expected to be exposed to local heat loads, depends significantly on the magnetic configuration, however some of the deposited power is due to energetic electrons, which have large poloidal orbits and are likely to be deposited on the plasma facing components.

DOI： 10.1088/1009-0630/18/11/03

Language：English   Publishing type：Research paper (scientific journal)  

In the present RF-driven divertor plasma of QUEST, it has been observed that orbit-driven current flows in the open magnetic surfaces outside of the closed magnetic surfaces. To observe this phenomenon and the characteristics of the orbit-driven current, current density profiles have been calculated on two different equilibrium conditions. We calculated current density profiles from particle guiding center orbits both for the fundamental and the second harmonic resonances for the 8.2 GHz electron cyclotron current drive. From this calculation, hollow current density profiles have been obtained with significant characteristics on both conditions. Only positive current distribution has been observed in the open magnetic surfaces outside of the closed magnetic surfaces.

DOI： 10.1016/j.fusengdes.2015.11.007

Language：English   Publishing type：Research paper (scientific journal)  

As QUEST has a small central solenoid (CS), a larger Ohmic discharge current has been obtained when the plasma shifts to the inboard side. This tendency restricts a divertor operation to the smaller plasma current regime. As the inner divertor coil has a smaller mutual inductance, it would be expected that its utilization seems to be better for easier plasma current ramp-up for a divertor operation. In this work, we made comparative studies on the plasma current ramp-up for two divertor coils. It is found that while the inner divertor coil with smaller mutual inductance needs a larger coil current, the outer divertor coil with larger mutual inductance needs a smaller coil current for divertor operation. Thus we have found that the plasma current ramp-up characteristics are almost similar for both configurations. We also propose a new fueling method for spherical tokamak (ST) using the coaxial helicity injection (CHI). The main plasma current would be generated at first, and then the CHI plasma current is created between bottom two electrode plates and merged into the main plasma current for fueling.

DOI： 10.1016/j.fusengdes.2015.12.022

Language：English   Publishing type：Research paper (scientific journal)  

In the present RF-driven (ECCD) steady-state plasma on QUEST (Bt = 0.25 T, R = 0.68 m, a = 0.40 m), plasma current seems to flow in the open magnetic surface outside of the closed magnetic surface in the low-field region according to plasma current fitting (PCF) method. We consider that the current in the open magnetic surface is due to orbit-driven current by high-energy particles in RF-driven plasma. So based on the analysis of current density profile based on the orbit-driven current, plasma equilibrium is to be calculated. We calculated high energy particles guiding center orbits as a contour plot of conserved variable in Hamiltonian formulation and considered particles initial position with different levels of energy and pitch angles that satisfy resonance condition. Then the profile of orbit-driven current is estimated by multiplying the particle density on the resonance surface and the velocity on the orbits. This analysis shows negative current near the magnetic axis and hollow current profile is expected even if pressure driven current is considered. Considering the hollow current profile shifted toward the low-field region, the equilibrium is fitted by J-EFIT coded by MATLAB.

DOI： 10.1016/j.fusengdes.2015.11.035

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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 Γinc is numerically reconstructed by fitting the experimentally measured hydrogen flux Γp, 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% of the Γinc modulation, or ∼1017 H m−2s−1.

DOI： 10.1016/j.vacuum.2016.04.025

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

DOI： 10.1063/1.4928878

Language：English   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/j.jnucmat.2014.12.092

Language：English   Publishing type：Research paper (scientific journal)  

DOI： 10.1016/j.jnucmat.2014.11.136

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Progression from low (LR) to high recycling (HR) was observed in full non-inductive long duration discharges up to 12 minutes on QUEST. Repetitive transition between LR and HR was induced by periodic gas puffing and the period to recover to LR, rec, was gradually prolonged with the plasma duration. The period, rec, normalized by gas rate has a linear relation to time-integrated H. This indicates hydrogen recycling induced an intentional gas puffing is increasing with time-integrated H which an indicator of hydrogen (H) fluence to the wall. To understand the dependence, deuterium (D) storing capability of the specimen exposed to QUEST plasmas was investigated by implantation of D2+
ions of 1keV and subsequent thermal desorption spectroscopy (TDS) as a post-mortem analysis. A diffusion-trap model, reconstructing the obtained the results of TDS, shows a clear dependence of recycling ratio on fluence and a time evolution of wall-stored H can be reconstructed. The model calculation indicates surface recombination on the deposition layer plays an essential role in QUEST wall behaviour. Wall temperature dependence on the number of injected H to keep density is experimentally investigated and is agreed with the model calculation. This result is quite helpful to execute hot wall experiments on QUEST.

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

In multi-pass Thomson scattering (TS) scheme, a laser pulse makes multiple round trips through the plasma, and the effective laser energy is enhanced, and we can increase the signal-to-noise ratio as a result. We have developed a coaxial optical cavity in which a laser pulse is confined, and we performed TS measurements using the coaxial cavity in tokamak plasmas for the first time. In the optical cavity, the laser energy attenuation was approximately 30% in each round trip, and we achieved a photon number gain of about 3 compared with that obtained in the first round trip. In addition, the temperature measurement accuracy was improved by accumulating the first three round trip waveforms.

Language：English   Publishing type：Research paper (scientific journal)  

In multi-pass Thomson scattering (TS) scheme, a laser pulse makes multiple round trips through the plasma, and the effective laser energy is enhanced, and we can increase the signal-to-noise ratio as a result. We have developed a coaxial optical cavity in which a laser pulse is confined, and we performed TS measurements using the coaxial cavity in tokamak plasmas for the first time. In the optical cavity, the laser energy attenuation was approximately 30% in each round trip, and we achieved a photon number gain of about 3 compared with that obtained in the first round trip. In addition, the temperature measurement accuracy was improved by accumulating the first three round trip waveforms. (C) 2014 AIP Publishing LLC.

DOI： 10.1063/1.4878260

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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% 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

Language：English   Publishing type：Research paper (scientific journal)  

A multi-pass Thomson scattering (TS) system based on confining laser pulses in an optical cavity was constructed for measuring very low-density plasma in the TST-2 spherical tokamak device. This paper describes the setup of the optical system, injection of the laser pulse into the cavity, and properties of the confined laser pulse. A combination of Pockels cell plus polarizer, which serves as an optical shutter, allows us to inject and then confine intense laser pulses in the cavity. A photodiode signal monitoring the very weak light leaking from the cavity mirrors demonstrated that the laser pulse makes many round trips, with a round-trip efficiency of approximately 0.73. The effective number of round trips (i.e., the signal enhancement factor) is approximately 3.7. For an injection efficiency of approximately 0.69, a cavity-confined laser pulse, applied to Thomson scattering, will yield a scattered signal that is five times larger than that from a single-pass laser pulse.

Language：English   Publishing type：Research paper (scientific journal)  

Language：Japanese   Publishing type：Research paper (other academic)  

Language：English   Publishing type：Research paper (other academic)  

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：Japanese   Publishing type：Research paper (other academic)  

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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% of the time, can carry more than 60% loss of the entire radial particle flux. (C) 2013 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2013.01.106

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Fusion virtual laboratory (FVL) is the experiments' collaboration platform covering multiple fusion projects in Japan. Major Japanese fusion laboratories and universities are mutually connected through the dedicated virtual private network, named SNET, on SINET4. It has 3 different categories; (i) LHD remote participation, (ii) bilateral experiments' collaboration, and (iii) remote use of supercomputer. By extending the LABCOM data system developed at LHD, FVL supports (i) and (ii) so that it can deal with not only LHD data but also the data of two remote experiments: QUEST at Kyushu University and GAMMA10 at University of Tsukuba. FVL has applied the latest "cloud" technology for both data acquisition and storage architecture. It can provide us high availability and performance scalability of the whole system. With a well optimized TCP data transferring method, the unified data access platform for both experimental data and numerical computation results could become realistic on FVL. The FVL project will continue demonstrating the ITER-era international collaboration schemes and the necessary technology. (c) 2012 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2012.04.027

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (other academic)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (bulletin of university, research institution)  

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：Japanese   Publishing type：Research paper (other academic)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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。

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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% 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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (other academic)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

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.

Language：English   Publishing type：Research paper (scientific journal)  

A double-pass Thomson scattering system, in which a laser pulse makes a round trip through the plasma, was constructed. Using the same optics and a fast detection unit, we can resolve backward and forward scattering pulses in the signal. Because these scatterings reflect velocity distribution along different directions, electron temperature anisotropy can be estimated from the double-pass Thomson scattering system.

Language：English   Publishing type：Research paper (other academic)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

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.

Language：English   Publishing type：Research paper (bulletin of university, research institution)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：Japanese   Publishing type：Research paper (other academic)  

Language：English   Publishing type：Research paper (scientific journal)  

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).

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

A high-performance data acquisition system (LABCOM system) has been developed for steady state fusion experiments in Large Helical Device (LHD). It also acquires experimental data of multiple remote machines through the 1 Gbps fusion-dedicated virtual private network (SNET) in Japan. The LABCOM system consists of distributed data acquisition (DAQ) computers, data storage devices, and the index database of data locations. Its key objectives are (1) real-time DAQ with the same sampling rates as burst acquisition, and (2) scalability of both the performance and quantity, even in the number of sites. LHD now uses 80 DAQs in parallel and acquires 10.6 GB/shot raw data in short-pulse experiments of 3 min. iteration. In steady state operation, huge kHz or MHz fluctuation data for plasma diagnostics must be processed in real time. In LHD, each DAQ computer can deal with max. 110 MB/s continuous data. The DAQ cluster has established the world record of acquired data amount of 90 GB/shot, which almost reaches the ITER data estimate of 100 or 1000 GB/shot. As for the data storage, the massively parallel processing (MPP) structure is important for scalable input/output performance and data redundancy. Hundreds of tera-byte compressed data are stored in the two-stage storage devices. The data mediation service is a distinguishing characteristic that manages the peer-to-peer data handling among many server and client computers. As every element of the LABCOM system are ds every element of the LABCOM system are distributed on the local area network (LAN), the data of remote fusion devices are acquired simply by extending the LAN to the wide-area SNET. LABCOM system acquires data from three remote experiments of different universities. The speed lowering problem in distant TCP/IP communication, e.g. 60 Mbps on 1 Gbps SNET, is improved by using the optimized congestion control and packet pacing technology. Its bandwidth tests between Japan and France achieved effective 881 Mbps over 10 000 s. A light-weight access control is implemented in the LABCOM system because it provides enough security for a closed VPN like SNET. Toward the fusion goal, a common data-accessing platform is indispensable so that physicists can easily make detailed comparisons between multiple large and small experiments. The demonstrated bilateral scheme will be analogous to that of ITER and the supporting machines.

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

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.

Language：English   Publishing type：Research paper (international conference proceedings)  

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.

Language：English   Publishing type：Research paper (international conference proceedings)  

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： 0.0

Language：English   Publishing type：Research paper (scientific journal)  

Language：English  

Language：English  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

The data acquisition (DAQ) and management system of the Large Helical Device (LHD), named the LABCOM system, has been in development since 1995. The recently acquired data have grown to 7 gigabytes per shot, 10 times bigger than estimated before the experiment. In 2006 during 1-h pulse experiments, 90 gigabytes of data was acquired, a new world record. This data explosion has been enabled by the massively distributed processing architecture and the newly developed capability of real-time streaming acquisition. The former provides linear expandability since increasing the number of parallel DAQs avoids I/O bottlenecks. The latter improves the unit performance from 0.7 megabytes/s in conventional CA MAC digitizers to nonstop 110 megabytes/s in CompactPCI The technical goal of this system is to be able to handle one hundred 100 megabytes/s concurrent DAQs even for steady-state plasma diagnostics. This is similar to the data production rate of the next-generation experiments, such as ITER. The LABCOM storage has several hundred terabytes of storage in double-tier structure: The first consists of tens of hard drive arrays, and the second some Blu-ray Disc libraries. Multiplex and redundant storage servers are mandatory for higher availability and throughputs. They together serve sharable volumes on Red Hat GFS2 cluster file systems. The LABCOM system is used not only for LHD but also for the QUEST and GAMMA10 experiments, creating a new Fusion Virtual Laboratory remote participation environment that others can access regardless of their location.

DOI： 0.0

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Least square method has been used to solve the inverse problem from flux loop measurement to the boundary
values on the CCS (Cauchy Condition Surface). When the CCS method is applied to real experimental data, noise
superposition is inevitable. By introduction of SVD (Singular Value Decomposition) and truncation of the least SV
components, we can expect the shape reproduction robustness against the noise. The truncation simulation of the
less SV component shows shape reproduction error less than no truncation, when the measurement error exceeds a
certain value.

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：Japanese   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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 wave form. 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： 0.0

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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%) 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

Language：English   Publishing type：Research paper (international conference proceedings)  

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： 0.0

Language：English   Publishing type：Research paper (international conference proceedings)  

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： 0.0

Language：English   Publishing type：Research paper (bulletin of university, research institution)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

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% 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% 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.

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Compact toroid (CT) injection experiments have recently been carried out in a compact plasma wall interaction experimental device (CPD) in the presence of toroidal as well as poloidal magnetic field and low density RF target plasma. Spectroscopic studies have been carried out using an IR camera, an IR spectrometer and an H(alpha) monitor to estimate the heat load deposition on the target and also to study the magnetic field as well as RF target plasma effects on CT plasma penetration behaviour. Using IR camera imaging of the target plate, heat load deposition on it has been estimated from the observed temperature rise of the plate and has been compared with the energy content of CT plasma. The estimation suggests that nearly 700 J of energy is uniformly deposited on the target plate under vacuum condition without any magnetic field. On the other hand, post exposure analysis of target plate using scanning electron microscopy has indicated local melting of the target plate. It has been observed that radial penetration of CT towards the target plate is reduced with the increase in the external magnetic field, both toroidal and poloidal. Moreover, the presence of RF target plasma in external magnetic fields seems to change the CT penetration behaviour significantly.

DOI： 10.1088/0029-5515/48/10/105001

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (bulletin of university, research institution)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (scientific journal)  

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 = ±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.

Language：English   Publishing type：Research paper (scientific journal)  

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 various magnetic field configurations on RF plasma confinement topology. The performance of the lithium sheet beam is absolutely calibrated by a quartz crystal monitor. Experimental results show that plasma initiation takes place at the electron cyclotron resonance layer in a simple torus configuration and then it expands quickly to the low magnetic field side. Different magnetic field configurations critically affect the RF plasma confinement topology. A sharp lower boundary exists for the RF plasma in magnetic null configuration. Magnetic connection length plays the key role in defining plasma boundary and the critical value of connection length for plasma to exist in CPD is found to be ∼ 5-6 meter for a given pressure condition.

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

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： 0.0

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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： 0.0

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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% increment. (c) 2007 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2007.01.201

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

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.

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

As for hot gas flow according to fluid equation with interaction with electrical arc and solid chamber, fluid equation must be solved with moving boundary of the chamber and with interaction with hot arc and impurity intrusion. Here we propose application of CIP (Cubic Interpolated Propagation) scheme, which is convenient and good at efficient advection and boundary motion. The CIP code was tested in simulation of normal shock tube, which has an analytical solution according to Rankine-Hugoniot relation, and applied in radial shock tube. And it was tested also in simulation of shock tube with solid piston between two fluids, using density function and functional transformation.

DOI： 0.0

Language：English   Publishing type：Research paper (international conference proceedings)  

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： 0.0

Language：English   Publishing type：Research paper (international conference proceedings)  

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.

DOI： 0.0

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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% 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.

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

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.

DOI： 0.0

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

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.

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：English   Publishing type：Research paper (scientific journal)  

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

Language：Japanese   Publishing type：Research paper (bulletin of university, research institution)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (scientific journal)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Language：English   Publishing type：Research paper (international conference proceedings)  

Event date： 2022.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：Online meeting   Country：Japan  

Event date： 2022.2

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2021.11

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：遠隔形式   Country：Japan  

Event date： 2021.11

Language：English   Presentation type：Oral presentation (general)  

Venue：Online meeting   Country：Japan  

Event date： 2021.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：遠隔形式   Country：Japan  

Event date： 2021.7

Language：English   Presentation type：Oral presentation (general)  

Venue：Remote Participation: Culham Centre for Fusion Energy, Culham, United Kingdom   Country：Japan  

In order to generate high-performance plasma for future fusion power generation, it is desirable to keep high quality vacuum during experiment. Mass spectrometer is commonly used to monitor the vacuum quality and to record the amount of atoms and molecules in the vacuum vessel. Leak is the most serious accident to avoid that can nullify an experiment and even harm researchers. Detecting leaks are ever more important since it can be easily overlooked, e.g., when the deterioration in the vacuum degree is modest. This forces the researcher to carefully observe the vacuum and mass spectrometer data. This article presents a way to suggest potential leaks in the vacuum vessel by analyzing mass spectrometer data. This is done by utilizing the Euclidean distance between composition ratios at different times for the clustering using the daily composition ratio. We show that our cluster analysis is an effective way of separating these two cases, which results in a semi-automatic determination of leaks is more efficient than the current norm, which is to check many measures to find a small abnormality in the data manually. We plan further model improvements for long-term evaluation.

Other Link： https://conferences.iaea.org/event/244/

Event date： 2021.1

Language：English   Presentation type：Symposium, workshop panel (public)  

Venue：Web meeting   Country：Japan  

Extension of Operation Area for Steady State Operation on QUEST by Integrated Control with Hot Walls, The 29th International Toki Conference on Plasma and Fusion Research (ITC29)

Other Link： http://www.triam.kyushu-u.ac.jp/QUEST_HP/sub/workshop.html

Event date： 2020.10

Language：English   Presentation type：Oral presentation (invited, special)  

Venue：Ceratopia Toki, Toki-city, Gifu, Japan   Country：Japan  

Extension of Operation Area for Steady State Operation on QUEST by Integrated Control with Hot Walls, The 29th International Toki Conference on Plasma and Fusion Research (ITC29)

Other Link： https://itc.nifs.ac.jp/

Event date： 2018.7

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：核融合科学研究所 管理棟4階　第2議室   Country：Japan  

Event date： 2017.12

Language：English   Presentation type：Public lecture, seminar, tutorial, course, or other speech  

Venue：2F meeting room, QUEST Exp. building   Country：Japan  

●Introduction of QUEST
　・History of long duration discharges
●Exp. system including plasma control system on QUEST
●Usage of FPGA as software-defined technology
●Information sharing with Ethernet for coordinated operation
●Summary

Event date： 2017.9

Language：English   Presentation type：Oral presentation (general)  

Venue：Chengdu   Country：China  

Efforts toward Steady State Operation in Long Duration Discharges
with the Control of Hot Wall Temperature on QUEST

M. Hasegawa1, K. Hanada1, N. Yoshida1, A. Kuzmin1, H. Zushi1, K. Nakamura1,
A. Fujisawa1, H. Idei1, Y. Nagashima1, O. Watanabe1, T. Onchi1, H. Watanabe1,
K. Tokunaga1, A. Higashijima1, S. Kawasaki1, and T. Nagata1
1 RIAM, Kyushu University, Japan


Achievement of steady state operation (SSO) of magnetic fusion devices is one of important issues for fusion research. Fully non-inductive plasma start-up and its maintenance up to 1h55min was successfully achieved on QUEST with a microwave of 8.2GHz, 40kW and well-controlled gas fueling and plasma-facing wall (PFW) temperature of 373K. The gas fueling is feedback controlled to keep constant in H signal, which can be an indicator of in-coming H flux to plasma facing materials (PFMs). On QUEST, the hot wall, which can be actively heated by electrical heater, was installed inside the vacuum vessel in 2014 autumn/winter (A/W) campaign, and the plasma can be sustained with high temperature PFW to investigate particle balance such as fuel recycling and wall pumping properties. Thermal insulators are installed between hot wall and vacuum vessel wall to keep the temperature of vacuum vessel wall below 423K for the protection of various diagnostics and plasma-heating devices. The function of active cooling of hot wall with cooling water channels will be installed in 2017 spring/summer (S/S) campaign.
The plasma-wall interaction (PWI) is an important subject when considering SSO, and is a wide-range issue because the matters such as material science and the plasma science are linked each other complicatedly. In these matters, especially, power balance and particle balance play important roles against SSO. The power balance in long duration discharges was sufficiently investigated in TRIAM-1M, which has the world record of plasma duration on tokamaks for more than 5h16min [1]. During the long plasma discharge, all of the temperatures of PFMs are saturated and kept constant on TRIAM-1M. The power balance on QUEST is also investigated before 2014, in which the hot wall had been installed. Approximately 70%-90% of the injected power could be detected by calorimetric measurements of PFMs, and about half of the injected power was deposited on the vessel wall [2].
The total particle balance on QUEST is estimated experimentally [3]. The time evolution of wall-pumping rate is evaluated as the difference between injected and evacuated H2 flux, which are derived from the flowmeter installed on gas fueling system and a quadrupole mass analyzer (QMS) installed on the bottom of the vessel, respectively. Absolute values of them are calibrated with consideration of the pressure and volume of gas fueling line and the relationship between flowmeter and QMS signal with the situation of no plasma. The wall-stored H can be obtained by time-integration of wall-pumping rate with setting the initial integrated value at zero. On the QUEST, the wall kept at higher temperature is rather active, and almost all stored H particles are released from the wall during the intervals of plasma discharges.
In the long duration discharges, the wall pumping occurs in the initial phase, and its rate gradually decrease. Finally, the wall-pumping rate becomes zero, and the wall saturation occurs. This tendency is likely to occur faster when its wall temperature is higher. To express this tendency, a wall model with hydrogen barrier (HB) which is formed around boundary between the deposition layer and the substrate was proposed [4]. In this model, the time derivative of the number of H dissolved in wall (dHW/dt) is proportional to the square of HW, when the number of H trapped in defects (HT) can be negligible. The parabolic relation between dHW/dt and HW is clearly observed in low HW experimentally, and the given curves with this model is well-fitted to the experimental observation.
References
[1] H.Zushi, et al, Steady-state tokamak operation, ITB transition and sustainment and ECCD experiments in TRIAM-1M, Nuclear Fusion, 45 (2005) S142-S156
[2] K.Hanada, et al, Power Balance Estimation in Long Duration Discharge on QUEST, Plasma Science and Technology, 18 (2016) 1069-1075.
[3] K. Hanada, et al, Investigation of hydrogen recycling property and its control with hot wall in long duration discharges on QUEST, Nuclear Fusion, (2017) to be published.
[4] K. Hanada, et al, Particle balance in long duration RF driven plasmas on QUEST, Journal of Nuclear Materials, 463 (2015) 1084-1086.

Event date： 2017.7

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：九州大学応用力学研究所２階会議室   Country：Japan  

●QUESTにおける制御システム
●長時間用プラズマ同定と制御
●粒子供給フィードバック制御
●高温壁の温度制御とその他
●まとめ

Event date： 2017.6

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：C-cube, 九州大学筑紫キャンパス   Country：Japan  

QUESTにおける定常プラズマ実験に向けた制御

核融合力学部門
プラズマ表面相互作用分野
　長谷川真

トカマク装置など磁場閉じ込め式核融合装置において、統合的にプラズマを制御して定常運転を実現することは非常に重要な課題である。QUEST装置では、この課題解決に精力的に取り組んでおり、温度を制御することができる高温壁をプラズマ対向壁として有していることが装置の特徴の一つとして挙げられる。定常運転では、プラズマ対向壁の温度が非常に大きな役割を持ち、たとえば２００℃の壁の温度でのプラズマ放電においては、壁の粒子吸蔵量が増えてくると、１２０℃の場合に比べて壁排気量が下がることが実験的に示されている[1]。現在までにQUEST装置では壁の温度を１２０℃に保ったまま、４０kW程度の加熱入力パワーで１時間５５分に及ぶ長時間放電を実現しているが、今後更にプラズマの加熱入力パワーを増やして、主プラズマが高密度、高温度の領域においても本課題に取り組む予定である。
QUESTの高温壁には、壁の温度を高温に保つためのシースヒーターが内蔵されており、シーケンサーを用いて壁の温度をモニターしつつシースヒーターの消費電力を調整して温度をフィードバック制御する加熱システムが２０１４年に実装されている。一方、プラズマの加熱入力パワーが増えて壁への熱入力が増えてくると、加熱システムは、温度を一定に保とうとした場合、ヒーターの消費電力を下げてき、遂にはゼロにするが、更にプラズマの加熱入力パワーが増える場合には、逆に冷却する機構が必要になる。この冷却に求められる能力として、数十kW程度を見込んでおり、空気冷却では冷却能力が足りず、直接接触する水冷却では過大になると予想されることから、伝熱板を介して冷却管をパネルに接触させる間接冷却の機構が採用されている。本方式の実際の冷却能力等は今後調査していくことになるが、この冷却では電動バルブを開閉することで、冷却の有無を調節する。この調節には加熱システムや中央制御システム、またプラズマ制御システムなど、各機器間の連係動作が必要になる。講演では、これら加熱・冷却システムの紹介や、プラズマを長時間維持するための、いくつかの取り組みについて紹介する。

[1] 26th IAEA Fusion Energy Conference, EX/P4-49, K. Hanada, et. al.

Event date： 2017.5

Language：English   Presentation type：Oral presentation (general)  

Venue：Max-Planck-Institut für Plasmaphysik, Greifswald   Country：Germany  

Modifications of Plasma Control System and Central Control System for Integrated Control of Long Plasma Sustainment on QUEST

Makoto Hasegawa1, Kazuo Nakamura1, Kazuaki Hanada1, Shoji Kawasaki1, Arseniy Kuzmin1, Hiroshi Idei1, Kazutoshi Tokunaga1, Yoshihiko Nagashima1, Takumi Onchi1, Kengoh Kuroda1,
Osamu Watanabe1, Aki Higashijima1, and Takahiro Nagata1

1Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka, Japan
hasegawa@triam.kyushu-u.ac.jp, nakamura@triam.kyushu-u.ac.jp, hanada@triam.kyushu-u.ac.jp, kawasaki@triam.kyushu-u.ac.jp, kuzmin@triam.kyushu-u.ac.jp, idei@triam.kyushu-u.ac.jp, tokunaga@riam.kyushu-u.ac.jp, nagashima@triam.kyushu-u.ac.jp, onchi@triam.kyushu-u.ac.jp, kuroda@triam.kyushu-u.ac.jp, rwata@riam.kyushu-u.ac.jp, higashi@triam.kyushu-u.ac.jp, nagata@triam.kyushu-u.ac.jp

Achievement of steady state operation (SSO) is one of important issues for future magnetic fusion devices. The world record of plasma duration on tokamaks for more than 5h16min was achieved in TRIAM-1M [1], where particle balance and power balance are investigated. On QUEST, which is a middle sized spherical tokamak installed on the same place after the closing of TRIAM-1M experiments, these issues are also vigorously investigated, and the fully non-inductive plasma start-up and its maintenance up to 1h55min was successfully achieved [2] with a microwave of 8.2 GHz, 40 kW and well-controlled gas fueling and plasma facing wall (PFW) temperature of 373 K.
On QUEST, the hot wall which can be actively heated by electrical heaters was installed inside vacuum vessel in 2014, and the plasma discharge is sustained with high temperature PFW to investigate particle balance such as wall pumping properties. The function of active cooling for hot wall with the cooling water will be installed in 2017 spring. These controls of heating with electrical heaters and cooling with cooling water will be managed by the central control system and its peripheral subsystems with the coordination of them. On the other hand, the gas fueling during plasma discharge is feed-back controlled with referring to the signal level of H which is an indicator of in-coming H flux to PFWs. This control is managed by a Proportional-Integral-Differential (PID) control on the plasma control system using a mass-flow controller. The modifications and coordination of these control systems for long discharges are introduced.

1. H.Zushi, et al, “Steady-state tokamak operation, ITB transition and sustainment and ECCD experiments in TRIAM-1M”, Nuclear Fusion, 45 (2005) S142-S156.
2. K. Hanada, et al, Investigation of hydrogen recycling property and its control with hot wall in long duration discharges on QUEST, Nuclear Fusion, (2017) to be published.

Event date： 2017.2

Language：Japanese   Presentation type：Symposium, workshop panel (public)  

Venue：九州大学応用力学研究所   Country：Japan  

●QUESTにおける制御システム
●プラズマ制御システムの紹介
　・構成、仕様、機能など
　・FPGAの使用例
●中央制御システムの紹介
　・構成、現状の課題など
　・システムへの実装例の紹介
●まとめ

Event date： 2016.10

Language：Japanese   Presentation type：Symposium, workshop panel (public)  

Venue：核融合科学研究所（岐阜県土岐市）   Country：Japan  

QUEST装置の紹介
LabVIEW言語でのFPGA開発
FPGAの利用例
TFコイルの電気抵抗値監視
磁気計測
プラズマの電子密度計測
トリガー遅延分配器
まとめ

Event date： 2016.8 - 2018.8

Language：English   Presentation type：Oral presentation (general)  

Venue：C-Cube, Chikushi Campus, Kyushu University, Kasuga, Fukuoka, Japan   Country：Japan  

Event date： 2015.5

Language：English  

Venue：Nara Prefectural New Public Hall (Nara Kasugano International Forum "Iraka"), Nara   Country：Japan  

For long-pulse operation, the control of plasma current and its position is important to manage the heat loads, particle flux, and so on. Thus, the plasma current and its position have to be identified correctly during a long plasma discharge in real time. The plasma current and its position are usually calculated with signals of a rogowski coil sensor, magnetic flux loop sensors, and magnetic pick-up coil sensors. In order to calculate with these signals, the time-integrations of raw signals with electrical circuits or numerical calculations are required. However, these time-integrations cause drift errors which become larger according to the duration of the plasma discharge. And, this disturbs the correct identification and the control of a long-pulse plasma discharge.
We propose the use of hall sensors for the long-pulse operation. A hall sensor does not require time-integration, and does not cause the drift error. On QUEST, several hall sensors are installed on the outside of the vacuum vessel. Although the quick behavior of plasma cannot be sensed with hall sensors because of eddy current effects, this enables high-accuracy measurements with the static environment of no plasma, no RF, and no vacuum. This also enables the repair or replacement of hall sensors without a vacuum purge. The plasma current and its position are identified in real time with these hall sensor signals. The plasma position and current are calculated by the evaluation of intensity ratios and intensities itself, respectively.

Other Link： http://tmsso2015.nifs.ac.jp/index.html

Event date： 2015.4

Language：English  

Venue：Nirma University, Auditorium of Institute of Pharmacy, Sarkhej-Gandhinagar Highway, Post : Chandlodia, Via : Gota, Ahmedabad - 382 481. Gujarat   Country：India  

Plasma control system (PCS) on QUEST has been developed for the achievement of the steady-state sustainment of tokamak plasma. QUEST is a spherical tokamak [1], on which high temperature all metal vessel wall up to 500 K is planned for the steady-state operation under unity recycling ratio. Achievement of steady-state operation in tokamak plasma is one of a key issue to realize cost-effective fusion power plants. In the aim of this, many kind of controls are required such as plasma position and its shape control, particle balance control, and heat load control. Current status and prospect of PCS for steady-state operation on QUEST are described.
For the control of plasma position and its shape, these parameters have to be identified in real time and steadily. Though magnetic sensors of rogowski coils, pick-up coils, and flux loops are usually used for this identification, these sensors are not suitable for the long time measurements because drift error induced by time integration occurs. On QUEST, in addition to these sensors, hall sensors are used, which are suitable for the long time measurement because of no drift errors. Furthermore, hall sensors can be expected to have an ease of maintenance and high accuracy because these are located on the outside of vacuum vessel wall where is less noisy environment compared to the inside one. The plasma current and position are calculated with just hall sensor signals, assuming the plasma as a filament current located on the inside of vacuum vessel. In this procedure, the plasma position and plasma current are evaluated with ratios and intensities of hall sensor signals, respectively. In addition to this, plasma shape is also evaluated in real time with a shape identification method [2]. These procedures are applicable to the control of plasma position and its shape for steady-state operation.
For the control of particle balance, a fueling feed-back control is implemented, which is referring Ha signals instead of plasma density. The fueling gas is puffed when an actual Ha signal intensity is lower the target intensity, and the actual signal gradually comes close to the target signal with a setting of the pulse prohibited duration. The actual Ha signal is well controlled with this method on over 10 minutes plasma discharge. Other several approaches such as a distributing system for steady-state operation will be discussed.

1. K. Hanada, K. Sato, H. Zushi, K. Nakamura, M. Sakamoto, H. Idei, et al., “Steady-State Operation Scenario and the First Experimental Result on QUEST”, Plasma and Fusion Research, 5, S1007 (2010).
2. M. Hasegawa, K. Nakamura, H. Zushi, K. Hanada, a. Fujisawa, K. Matsuoka, et al., “Development of plasma control system for divertor configuration on QUEST”, Fusion Engineering and Design, 88, 1074–1077 (2013).

Event date： 2015.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：九州大学　応用力学研究所   Country：Japan  

QUEST装置における制御システムの現状と展望の紹介

Event date： 2014.9 - 2014.10

Language：English  

Venue：Kursaal Congress Center, Av. Zurriola 1, 20002 Donostia - San Sebastián   Country：Spain  

Plasma control systems for tokamak plasmas are required to make control signals in real-time with simultaneously acquiring various data and calculating meaningful physical quantities. Since the physical quantities and the control signals have relationship with each other, a centralized control system is principally desirable for the grasp of these parameters. However, the computational loads on the CPU of plasma control workstation (WS) become too large to build a highly integrated control system, because it makes difficult to execute in real-time. In actual, the CPU utilization of the WS for the spherical tokamak QUEST becomes almost full.
We propose to develop a decentralized control system. In this system, each control system has a reflective memory connected to each other with optical fibers, and shares various data via reflective memory. The good point of this system is to increase the CPU resource. Furthermore, the electrical insulation is ensured spontaneously. On the other hand, the synchronization accuracy between each system may become worse.
The GE cPCI-5565PIORC of National Instruments Corporation is used as the reflective memory, which has 256 Mbytes memory and 170Mbyte/sec transfer rate. The most popular data type to share is double-precision real type (DBL) which needs 8 bytes to represent. The actual data read or write time is measured. Especially, within the period of 4 kHz which is the period of WS, more than 1000 to 2000 DBLs can be read or write. This means about 50 Mbytes/sec transfer rate for the one directional data sharing. For the bidirectional data sharing, each system has to repeat the read-write procedure. This would take more time. In the presentation, we will introduce the actual implementation of the reflective memory to the decentralized control system and its performance.

Event date： 2013.11

Language：English   Presentation type：Symposium, workshop panel (public)  

Venue：Gyeongju Hilton Hotel, Gyeongju City   Country：Korea, Republic of  

A long time plasma sustainment is an important issue for the future nuclear fusion plasma. In QUEST (Q-shu university experiment with steady-state spherical tokamak), a steady state operation is also one of project objectives. Thus, the long time identification and its control of the plasma position and its shape are important for the steady state operation. However, the long time identification is difficult, as long as the integrated magnetic signals such as magnetic fluxes or magnetic fields are used because the integration errors, namely, drift errors occur and prevent the accurate identification.
The WS is composed of PXI systems of the National Instruments Corporation, which contains a controller module (2.26 GHz Intel Core 2 Quad processor, memory: 2 GBytes) based on a real-time operating system, one DIO module (16-channel digital input and output), and six FPGA modules (eight-channel analog input and output in each module). In the WS, the several tasks can be performed in parallel because a multi-quad-core processor is used in the controller module. One task is for the control of a DIO module and FPGA modules. Another task, referred to as a main loop, is for the calculation of control signals by the acquired data. These two tasks are performed at 4 kHz. In addition, the real-time equilibrium calculation and the plasma image analysis are executed in parallel on other cores, respectively. The calculation period of the image analysis will be several seconds. That is sufficient to correct the drifts of magnetic fluxes. In this presentation, the development status of this control system will be introduced.

Other Link： http://www.nfri.re.kr/apfa2013/index.php

Event date： 2013.2 - 2013.3

Language：English   Presentation type：Oral presentation (general)  

Venue：RIAM QUEST Building, Kyushu Univ.   Country：Japan  

Event date： 2013.1

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：核融合科学研究所　管理棟４階　第１会議室   Country：Japan  

リアルタイム用平衡計算コードの開発を行った。本コードの1回の計算時間はメッシュ数30×48の場合において2msec以内に収まると予想される。今後、この平衡計算コードの収束性や妥当性を検証するために、リミター配位やダイバータ配位など幾つかの放電パターンに対して計算を行い、実験データ及び他の平衡計算コードとの比較等を行うなどして、より詳細に吟味し、QUESTの制御システムに実装していく予定である。また、制御においては、現状はプラズマ内側エッジ位置のみの制御であるので、形状制御と呼ぶに十分な制御可能個所の個数を増やす予定である。また、プラズマ形状だけでなく、ストライクポイント等のリアルタイム同定を行い、熱負荷制御等に資する予定である。

Event date： 2012.9

Language：English  

Venue：Liège   Country：Belgium  

A plasma control system in order to sustain divertor configurations is developed on QUEST (Q-shu university experiment with steady-state spherical tokamak). Magnetic fluxes are numerically integrated by 100 kHz frequency with usage of FPGA (Field-Programmable Gate Array) modules, and transferred to a main calculation loop with 4 kHz. With these signals, plasma shapes are identified in real time with 2 kHz frequency 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 position controls are tested using PID (proportional–integral–derivative) control loops for target positions. Whereas the outside edge position can not be controlled by outer PF coil current, the inside edge position can be controlled by inner PF coil current.

Event date： 2011.9

Language：English   Presentation type：Symposium, workshop panel (public)  

Venue：National Institute for Fusion Science, Toki   Country：Japan  

In a similar way to other spherical tokamaks, the achievement of steady state operation with divertor configuration is one of important issues for the QUEST project. The control system for this has been developed in the QUEST. The identification of plasma position and its configuration is required for the control. One of the methods adopted in this control system is to adjust plasma shape parameters such as elongation and triangularity directly so that the calculated magnetic signals become the same values as measured ones. Although this method cannot be adopted if the plasma shape is complicated, one can expect that the time to calculate become short because there is no need to calculate values such as flux values at inside area of vacuum vessel but just installed positions of magnetic sensors. This calculation method has been installed into the control system of the QUEST which is composed of 4 CPU cores and Real Time-OS and operates main control loop with 4 kHz period. And, this calculation with 22 flux loop signals is finished within 1msec by using parallel processing technology.
The horizontal and vertical plasma positions are controlled by active coils called HCUL coils and PF26 coils, respectively with simple PID control method. The current of PF26 coils changes not only vertical magnetic field but n-index. Furthermore, the n-index also varies gradually in the process that plasma configuration changes from limiter configuration to divertor configuration. This change affects vertical control, and the appropriate PID gain values differ by each magnetic configuration. For this, the mechanism to regulate each gain values automatically according to the magnetic configuration will be also installed into the control system.

Event date： 2011.2

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Reports of utilization of SNET on QUEST experiment

Other Link： http://ccpuck.nifs.ac.jp/research/H22-SNET-FusionVL.html

Event date： 2010.2

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：NIFS研究1期棟 6F会議室 (TV会議システム併用)   Country：Japan  

Event date： 2009.12

Language：Japanese  

Venue：千里阪急ホテル   Country：Japan  

Event date： 2009.12

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：福岡会場　九州大学箱崎キャンパス　附属図書館視聴覚ホール   Country：Japan  

Event date： 2009.2

Presentation type：Oral presentation (general)  

Venue：核融合科学研究所   Country：Japan  

Event date： 2008.9

Presentation type：Oral presentation (general)  

Venue：福岡国際会議場   Country：Japan  

Event date： 2008.7

Presentation type：Oral presentation (general)  

Venue：九州大学筑紫地区   Country：Japan  

Event date： 2008.2

Presentation type：Oral presentation (general)  

Venue：核融合科学研究所   Country：Japan  

Event date： 2008.2

Presentation type：Oral presentation (general)  

Venue：核融合科学研究所   Country：Japan  

Event date： 2007.6

Venue：Inuyama   Country：Japan  

Venue：筑波大学大学会館   Country：Japan  

Divertor design study using SOLDOR divertor simulation code on QUEST

Presentation type：Oral presentation (general)  

Venue：九州大学   Country：Japan  

Presentation type：Oral presentation (general)  

Venue：GA, San Diego   Country：United States  

Presentation type：Oral presentation (general)  

Venue：徳島文理大学   Country：Japan  

Presentation type：Oral presentation (general)  

Venue：Long Beach, CA   Country：United States  

Presentation type：Oral presentation (general)  

Venue：クローバープラザ（福岡県春日市）   Country：Japan  

Presentation type：Oral presentation (general)  

Venue：九州大学筑紫地区   Country：Japan  

Presentation type：Oral presentation (general)  

Venue：九州大学筑紫地区   Country：Japan  

Presentation type：Oral presentation (general)  

Venue：中部大学   Country：Japan  

Venue：核融合科学研究所   Country：Japan  

Presentation type：Oral presentation (general)  

Venue：核融合科学研究所(NIFS)   Country：Japan  

Presentation type：Oral presentation (general)  

Venue：九州大学応用力学研究所   Country：Japan  

Event date： 2008.6

Presentation type：Oral presentation (general)  

Venue：青森市男女共同参画プラザ（アウガ５F）   Country：Japan  

大型実験装置QUESTにおける真空排気設備のソフトウェア化とネットワーク化を行う。ソフトウェア化では、実際の設備をよく表した操作パネルや操作スイッチをディスプレイ上に表現した。ネットワーク化では機器間の通信をEthernetで行うことで多種多様なデータが機器間で共有できるようになった。これにより、拡張や変更等に柔軟に対応できるようになった。

大型実験装置QUESTにおける真空排気設備のソフトウェア化とネットワーク化を行う。ソフトウェア化では、実際の設備をよく表した操作パネルや操作スイッチをディスプレイ上に表現した。ネットワーク化では機器間の通信をEthernetで行うことで多種多様なデータが機器間で共有できるようになった。これにより、拡張や変更等に柔軟に対応できるようになった。

researchmap

プラズマを生成する実験設備では、安全上、プラズマ生成中は実験設備本体室の入室が禁じられている。本体室に人がいる状態で、誤ってプラズマを生成しないように警報音等の各種安全装置を設けているが、本入室管理システムもその一環に当たる。複数個所の出入り口に設置することが可能で、WiFi付ポケットサイズPCとタッチパネルを用いて省力で設置できるようにしてある。入退室時に自身の名前のボタンをタッチパネルでタッチする。中央制御装置と連動して入室者がある場合には、プラズマ生成のシーケンスは停止される。

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従来、QUEST装置で用いられたプラズマ形状のリアルタイム同定の手法は、プラズマ電流を一本のフィラメント電流と仮定して、所定の位置にフィラメントを置いたときの磁束分布の計算から最外殻磁気面を算出していた。しかし実際は、プラズマ電流は分布を持つものであり、実際の状態を再現しているとは言いがたい。そこで、Grad-Shafranov方程式と呼ばれる、プラズマの力学平衡の式をリアルタイムに解いて、プラズマ位置・形状を算出するコードの開発を行った。この計算方式の特徴としては、一回のデータ入力に対して、Grad-Shafranovの式を一回計算していることにある。すなわち、一回のデータ入力に対して、イタレーション回数が１回であり、収束性を確認しないでいる。この場合、同一のデータセットが何回(～数回)も入力された場合に計算結果が収束することになる。また、本計算の際に、他の方式で計算されたプラズマ位置をプラズマ位置の束縛条件として利用することができる。これは、プラズマが極度に非等方であり、等方を仮定している平衡計算が適切ではない場合に有効であり、すくなくとも磁気計測の計測値と計算値はほぼ一致する状態での計算結果を提供する。今後、この計算により求めた最外殻磁気面と多計測による位置の比較の検証を行い、また、最外殻磁気面の制御を行うよう低である。

トカマクプラズマの定常運転で問題となる磁気計測のドリフト問題の解決を試みる。ここでは、プラズマ画像からプラズマの位置・形状を同定して、それを元に計測されるであろう磁気計測の値を推定し、磁気計測の値に補正をかける。現状は、テスト動作での確認であるが、正常に動作することを確認している。
従来の磁気計測は、生信号を時間積分するため、積分時間が長くなればなるほど積分ドリフトと呼ばれるエラー量が発生していた。このため、磁気計測の結果から求められるプラズマの位置・形状を長時間にわたって同定することは難しかった。エラー量を軽減するために積分器の改良が試みられ、一定の効果が得られたものの、一度、周囲環境の温度変化やそれに伴う電子部品の特性変化などにより発生したエラー量が計測結果に混ざると、それを取り除くのは原理的に不可能である。そこで、本作品ではプラズマの画像を撮影し、その画像からプラズマの位置・形状を取得し、その情報から得られるであろう磁気計測の値を推定し、磁気計測に補正をかけることを試みている。プラズマの位置の同定方法は、現状は、グレースケールで撮影した映像に対して、ある値を境に白と黒の２値化を行い、明るい領域の図形重心を取るという方法を採用している。画像フィルターはNational Instruments Corp.より提供されているものを使用している。現状は、テスト動作を試したのみであるが、プラズマの明かりの代わりに懐中電灯を用いた試験では、プラズマの位置を同定し、その結果を下に線形に磁気計測値を補正することに成功している。実際のプラズマの位置・形状においても、National Instruments Corp.より多くの画像フィルターが提供されているので、これらを組み合わせることで同定できるものと考える。

トカマクプラズマの定常運転で問題となる磁気計測のドリフト問題の解決を試みる。ここでは、プラズマ画像からプラズマの位置・形状を同定して、それを元に計測されるであろう磁気計測の値を推定し、磁気計測の値に補正をかける。現状は、テスト動作での確認であるが、正常に動作することを確認している。

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従来、QUEST装置で用いられたプラズマ形状のリアルタイム同定の手法は、プラズマ電流を一本のフィラメント電流と仮定して、所定の位置にフィラメントを置いたときの磁束分布の計算から最外殻磁気面を算出していた。しかし実際は、プラズマ電流は分布を持つものであり、実際の状態を再現しているとは言いがたい。そこで、Grad-Shafranov方程式と呼ばれる、プラズマの力学平衡の式をリアルタイムに解いて、プラズマ位置・形状を算出するコードの開発を行った。この計算方式の特徴としては、一回のデータ入力に対して、Grad-Shafranovの式を一回計算していることにある。すなわち、一回のデータ入力に対して、イタレーション回数が１回であり、収束性を確認しないでいる。この場合、同一のデータセットが何回(～数回)も入力された場合に計算結果が収束することになる。また、本計算の際に、他の方式で計算されたプラズマ位置をプラズマ位置の束縛条件として利用することができる。これは、プラズマが極度に非等方であり、等方を仮定している平衡計算が適切ではない場合に有効であり、すくなくとも磁気計測の計測値と計算値はほぼ一致する状態での計算結果を提供する。今後、この計算により求めた最外殻磁気面と多計測による位置の比較の検証を行い、また、最外殻磁気面の制御を行うよう低である。

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トカマク型プラズマにおいて、プラズマ放電の長時間維持は重要な課題である。長時間維持を行うためには、プラズマを真空容器に接しないで維持させることが重要であると言われている。このためには、プラズマの形状をリアルタイムに同定して、その制御を行うことが求められる。本作品は、プラズマ形状のリアルタイム同定を行い、そのプラズマの境界位置を外部磁場コイルで制御するものであり、その形状の一部を本作品にて制御できることを確認した。
プラズマの形状を同定するには、プラズマからの応答である磁気の計測をプラズマ周辺で多く行うことが必要になる。このため、プラズマ制御システムの入出力チャンネル数をそれぞれ４８chから６４chへと増設した。チャンネル数の増設に伴い、サンプリング周波数等の処理速度の低下が予想されるが、DMA(Direct Memory Access)の利用や、並行実行となるコーディングの開発により、従来のサンプリング周波数を落とすことなく増設が行えている。
プラズマ形状の同定について、本作品ではプラズマ電流を一本のフィラメントと仮定して、その磁束分布を計算して最外殻磁気面位置を算出する方法を採用している。このようにすることで、形状位置の正確性はあまり保証されないが、短時間でプラズマ境界位置を計算できるという利点がある。この方法で、リアルタイムに計算することを可能にしている。
次に、一般的な形状制御ではフラックス値を参照して、そのフラックス値を制御するという方法が広く用いられている。ただ、この方法ではプラズマ境界位置とフラックス値を関係づけるのが難しいという欠点を有している。一方、本作品ではプラズマ境界位置を求め、その制御を直接行うことにしているので、より直観的な制御を可能としている。

トカマク型プラズマにおいて、プラズマ放電の長時間維持は重要な課題である。長時間維持を行うためには、プラズマを真空容器に接しないで維持させることが重要であると言われている。このためには、プラズマの形状をリアルタイムに同定して、その制御を行うことが求められる。本作品は、プラズマ形状のリアルタイム同定を行い、そのプラズマの境界位置を外部磁場コイルで制御するものであり、その形状の一部を本作品にて制御できることを確認した。

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高温プラズマ力学研究センターにおけるネットワーク通信において、九州大学のセキュリティーポリシーで禁止されているファイル交換ソフトのP2P通信を監視し、検出時にはその通信を遮断する仕組みの構築を行う。また、未登録端末については、HTTPおよびHTTPS通信といった基本的通信のみを許可し、全体的にセキュリティーレベルを向上させることを目的としたルータを構築した。
九州大学のセキュリティーポリシーで禁止されているファイル交換ソフトのP2P通信は、九大情報基盤センターでも監視しているが、かなり上流にある箇所での監視であり、P2P通信をおこなった特定の端末を探し出すことが難しい。特定の端末を探し出すためには、その端末に近い個所にあるルータで監視を行ったほうが効率的である。このため、高温プラズマ力学研究センター内にP2P通信を監視し、検出時にはその通信を遮断するルータの構築を行った。具体的には、2枚のネットワークボート(NIC)が挿入されたCentOS Linux端末を用意し、2枚のNICをブリッジ接続して透過接続を構成する。P2P通信の検出は無償で提供される「snort」によって行われる。ログ解析ツールである「swatch」を使用し、snortがP2P通信を検出したことをトリガーとして「iptables」を用いて該当IPアドレスの通信を遮断し、同時にネットワーク管理者に遮断が起きたことを電子メールで知らせている。ネットワーク管理者は、該当IPアドレスからMACアドレスを求め、該当端末を見つけだすことが可能となる。
また、センター内でのネットワーク通信では、外部研究者も自身の端末でネットワーク通信ができるように未登録端末の使用を認めている。ただし、この際には、HTTPおよびHTTPS通信といった基本的通信のみを許可するようにしている。端末を登録すると、ほぼすべてのポートを使用して外部と通信ができるようになる。このようにすることで、外部研究者の意図せぬファイル交換ソフトの使用を極力防ぎ、かつ登録端末であれば使用者への迅速な連絡が可能となる。
以上のようなルータを構築することで、センター内のネットワークセキュリティーの向上に貢献している。

高温プラズマ力学研究センターにおけるネットワーク通信において、九州大学のセキュリティーポリシーで禁止されているファイル交換ソフトのP2P通信を監視し、検出時にはその通信を遮断する仕組みの構築を行う。また、未登録端末については、HTTPおよびHTTPS通信といった基本的通信のみを許可し、全体的にセキュリティーレベルを向上させることを目的としたルータを構築した。

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Monitoring system of experiment hall entering personnel
Many personnel participate in the QUEST experiment. Because the high RF power is used and hard x-ray is generated during plasma sustain, the experiment operator is required to confirm that nobody stays in the experimental hall before the start of plasma discharge. Thus, the instrument which enables personnel to switch on the lamps of his name in the experimental hall is installed at the entrance of this hall, and this can communicate with the central control system to send the information of entering personnel. If any one stays in the hall, the central control system stops the experiment sequence according to the information of this system. This contributes for the safety operation of the QUEST.

QUEST装置を使用した実験は、同時に多人数の人間が参加する。一方、プラズマ生成時には大電力高周波を使用し、また硬X線も発生するので、本体室に入室者がいないことを確認して、プラズマの生成を行なう必要がある。そのため、入室者の有無を表示する機器を本体室入り口に設置し、その情報を中央制御に伝えるシステムを構築した。万一、入室者がいた場合には実験シーケンスは自動的に停止するようにしてあり、安全に実験が遂行できる。

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Central Control System of QUEST

球状トカマク装置QUESTにおける実験で、電源や高周波発信器など多くのサブシステムを統括する中央制御システムの継続的開発

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QUEST実験ではプラズマ閉じ込めのための磁場生成として、幾つかの電源とコイルを用いている。プラズマ放電を行なった際に使用した電源及びその接続コイルとコイルターン数などの結線の情報は、後解析を行なう際に決して欠くことのできない情報である。そこで、これらの情報を実験者はWEB上に記入して記録できるようにした。その記録の閲覧も、WEBで行なうことが可能である。また、中央制御システムと連動し、少なくとも、その日の実験の最初に記録を行なわないと、実験シーケンスの起動が行なえないようにしている。このようにすることで、記録を忘れることを未然に防ぎ、確実に正確な情報を記録できるようになった。

Other Link： http://www2.triam.kyushu-u.ac.jp/coilConfLog/

QUEST実験ではプラズマ閉じ込めのための磁場生成として、幾つかの電源とコイルを用いている。プラズマ放電を行なった際に使用した電源及びその接続コイルとコイルターン数などの結線の情報は、後解析を行なう際に決して欠くことのできない情報である。そこで、これらの情報を実験者はWEB上に記入して記録できるようにした。その記録の閲覧も、WEBで行なうことが可能である。また、中央制御システムと連動し、少なくとも、その日の実験の最初に記録を行なわないと、実験シーケンスの起動が行なえないようにしている。このようにすることで、記録を忘れることを未然に防ぎ、確実に正確な情報を記録できるようになった。

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We built and released the site which enables simple numerical calcuration such as magnetic configuration and electron trajectory on QUEST device.

Other Link： http://sim.triam.kyushu-u.ac.jp/simQUEST/

We built up and released the community site to all the collaborators for interactively sharing information such as technical, conference presentation, and research papers to promote studies and experiments of QUEST project.

Other Link： https://www.triam.kyushu-u.ac.jp/community/

QUEST装置を用いた実験・研究において、全国の共同研究者と技術情報や学会発表、論文等の情報の双方向の共有を可能にし、QUEST実験・研究の効率化・円滑化を目的とするサイトの構築を行なう。

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QUESTの磁場構造や電子軌道などの数値計算をWEB上で行ないグラフ表示するようにしたサイト。プラズマの力学的平衡(リミタ配位、ダイバータ配位)の計算も可能であり、計算結果をContour図などで視覚的に表現する。これらの計算において使用するコイルもWEB上にて選択が可能であるので、柔軟に実験状況に即した計算が可能となり、実験パラメータの選定に指針を与え、効率的な実験・研究の推進に貢献している。

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QUEST装置で生成されるプラズマを集中して制御するためのシステム。プラズマの応答を計測し、それに対応して各コイル電源、RF発振器、ガス供給等装置などへ指令値を出力して制御するシステム。
QUEST装置で生成されるプラズマを集中して制御するためのシステム。各コイル電源、RF発振器、ガス供給等装置などへ指令値を出力して制御する。また、リアルタイムでプラズマ位置の同定を行い、そのフィードバック制御を行ったり、プラズマ電流をコントロールするためのRFフィードバック制御、周回電圧フィードバック制御など、多種のフィードバック制御を実装し、適宜改良し、または最適なフィードバックループの開発を行っている。今後更に、プラズマ形状制御などのフィードバック制御を実装し、また幾つかのアクチュエータとセンサー(ex. 壁ヒーター 壁温度、クライオポンプ 排気量など)を追加しダイバータ配位の定常運転に向けた改造を行なう。

QUEST装置で生成されるプラズマを集中して制御するためのシステム。プラズマの応答を計測し、それに対応して各コイル電源、RF発振器、ガス供給等装置などへ指令値を出力して制御するシステム。

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小型PWI実験装置CPDにおいて、プラズマを生成、制御するためのソフトウェアであり、４つの磁場コイルのPID制御を行い、ガスパフ、高周波の入射を高精度のタイミングで制御を行う。

小型PWI実験装置CPDにおいて、プラズマを生成、制御するためのソフトウェアであり、４つの磁場コイルのPID制御を行い、ガスパフ、高周波の入射を高精度のタイミングで制御を行う。

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プラズマ放電を行う際の種々のパラメータをWeb上にて入力し、履歴を保管するデータベースであり、Web上にて履歴の確認や、パラメータの検索等をおこない、実験の効率化を諮るデータベースである。

プラズマ放電を行う際の種々のパラメータをWeb上にて入力し、履歴を保管するデータベースであり、Web上にて履歴の確認や、パラメータの検索等をおこない、実験の効率化を諮るデータベースである。

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プラズマ実験において収集した計測データを系統的に保管管理し、Web上にて利用者からの要求に応じてデータをグラフ化して表示、データのダウンロード、データの抽出などを行う。

プラズマ実験において収集した計測データを系統的に保管管理し、Web上にて利用者からの要求に応じてデータをグラフ化して表示、データのダウンロード、データの抽出などを行う。

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