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

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.

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

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 heating (EBWH) 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 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 EBWH 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 EBWH wave. Remarkably, a high hard-X-ray-radiation temperature ~ 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.

In this paper, we proposed the mathematical model of campus by measurement data and theoretical approach. We measure dynamic characteristic and static characteristic of load in a system. Power capacitors in the university are assumed to be electric power customer's load. The load model is presumed by combining them and the linear model calculated theoretically. The line voltage and the line current are measured, and we calculate the line load capacity in distribution system of campus. The load fluctuation of distribution system during a day is observed. The load will be modelized by evaluating a lot of data. We can forecast the demand of electric power by few information that is the load model and the measurement data of receiving end.

Recently, the number of system interconnection of the renewable energy sources (RES) such as the photovoltaic generation (PV) and wind power generation is increasing drastically, and there is in danger of changing the voltages in a distribution system by the precipitous output variation of RESs. In this study, the authors propose one voltage control method of the distribution system by the power factor control of plural PV systems in consideration of cooperation with the load ratio control transformer (LRT) of laggard control response installed beforehand in the distribution system. In the proposed method, the slow voltage variation is controlled by LRT, and the steep voltage variation uncontrollable by LRT is controlled by plural PV systems, as a result, all the node voltages are controllable within the proper limits. In order to verify the validity of the proposed method, the numerical calculations are carried out by using an analytical model of distribution system which interconnected PV systems.

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 (PDP). This permeation may also be significant, even in the walls, which are not directly exposed to the plasma. Under similar conditions, the permeation flux (Γperm) of hydrogen through a 30 μm thick Ni membrane heated at 412–575 K has been measured in the spherical tokamak QUEST. Γperm is being measured during the scans of different operating parameters like RF power (PRF), chamber pressure (Pchamber), discharge widths (τdis) and vertical magnetic field (BZ). 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 Γperm i.e. permeation fluence (Qperm) and the time integrated Hα intensity i.e. Hα fluence (Qα). Qperm 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.

In recent years, the number of connection of distributed generators (DGs) such as the photovoltaic generation (PV) and wind power generation is increasing, and there is in danger of changing the voltages in a distribution system by the precipitous output variation of DGs. In this study, the authors propose one voltage control method of a distribution system by the static var compensator (SVC) in consideration of cooperation with the load ratio control transformer (LRT) of laggard control response installed beforehand in the distribution system. The proposed method may be able to make the rated capacity of SVC small and may be able to reduce the aggravation of power factor of the distribution system, by setting up the dead band of voltage control appropriately. And the authors propose one determination method of the necessary minimum rated capacity and optimum control parameters of SVC in view of the cost of equipment and high-speed controllability. In order to verify the validity of the proposed method, the numerical calculations are carried out by using a distribution system model.

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 m in the case where a null point is set to be R=0.56 m 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.

A discharge longer than 5 h 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.

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.

An electron cyclotron heating (ECH) system has been installed on the TRIAM-1M tokamak, which produces a 170 GHz microwave up to 260 kW at the window of the gyrotron for 5 sec. After the Matching Optical Unit (MOU) converts the microwave to a Gaussian beam, a microwave power of 200 kW is transmitted by a corrugated waveguide of 63.5 mm in diameter. Tests confirmed these specifications, and the fundamental ordinary electron cyclotron (EC) wave of 170 GHz up to 120 kW was injected into the vacuum vessel of the TRIAM-1M in order to ionize neutral hydrogen gas to make source plasma, which is expected to reduce flux consumption in the case of current start-up by means of Ohmic heating (OH) coils. Though the breakdown does not occur at the input RF power, PRF = 40 kW with the neutral gas pressure, p = 1.3×10−3Pa, high-density plasma (more than 0.7×1019m−3) is produced at PRF = 100 kW. In the range of PRF = 40−100 kW, the electron density is proportional to PRF.

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.

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.

A double piston-actuated structure designed for replacing the use of diaphragms in conventional shock tubes has been recently developed. In order to clarify the piston sliding process for this structure, a simplified model based on the motion equation of its main sliding piston has been developed. Calculated piston sliding time against pressure ratio has been numerically obtained and indirectly evaluated by examining the experimental curves of the performance of the diaphragmless structure.

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.

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).

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.

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.

A large diameter diaphragmless shock tube has been recently developed and designed to perform detailed studies of CO2-N2 gas-dynamic laser (GDL). This large diameter diaphragmless shock tube offers various advantages over the conventional shock tubes (diaphragm-type) as longer test times, higher degree of reproducibility of shock-tube data, and especially low-impurity operation condition. The latter advantage is experimentally demonstrated herein, which is very critical issue in the CO2-N2 GDL studies. A supersonic nozzle section was mounted at the end wall of the shock tube and instrumented for simultaneous measurement of laser output power and energy. The GDL action in a CO2-N2 mixture under low impurity condition has been obtained by using the large diameter diaphragmless shock tube for the first time.

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.

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

A method of estimating the plasma elongation and position using the magnetic field picked up by magnetic coils is described. The method is based on the approximate solution for the Grad-Shafranov equation including the effect of the eddy current on the resistive shell. When this method was applied to noncircular plasmas on TRIAM-1M, the analytical results were found to be in good agreement with the plasma shape observed by a TV camera and with the outermost magnetic surface obtained through the calculation of an MHD equilibrium code. This approach is suitable for the real-time analysis and feedback control of plasma shape in a noncircular tokamak.