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Magnetic interaction between a tokamak reactor and its reinforced-concrete building has been analyzed using the analytical method and ANSYS electromagnetic code. Although it is better to use the non-magnetized material for the foundation of the air core tokamak building, it is interesting to study whether the magnetic material can be used in order to save the total cost of a fusion reactor. This study is motivated from the present ITER under construction in France. Assuming the iron plate approximation instead of discrete rebars, we analyzed the effect of the magnetized material on the magnetic field null regime in the plasma break down phase, x point location in the divertor operation, the magnetic force applied to the floor, and the induction effect on the iron floor. Although the effect of the magnetic material on the plasma performance is found to be not so crucial, the mechanical soundness of the floor due to the magnetic attractive force is most concern for machine safety.

With respect to a particle fueling method in a spherical tokamak reactor with an advanced fuel, we investigate the availability of a merging process following the translation of a small torus plasma through which a plasma current flows in the same direction as the main plasma. A three-dimensional resistive MHD model is used to examine the particle fueling effect by reproducing the translation process and simultaneously performing an ion trajectory calculation. The simulation shows the merging process between the translated plasma and the main plasma and exhibits potential as a particle fueling method. The results of the study indicate that up to 30% of the ions in the secondary plasma are supplied to the main plasma.

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.

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.

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

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.

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.

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.

Plasma operation control of the Large Helical Device (LHD)-type helical fusion reactor, FFHR-d1, was examined by using a quasi-one-dimensional particle balance calculation code. It was found that feedback control of the fusion power, which is proposed in the past study with a zero-dimensional model, was difficult because of the time delay between the pellet injection and response of the fusion power due to shallow pellet deposition. While the achievable parameter range of the fusion power at a steady state is limited by the pellet injection condition and magnetic field strength, smooth ignition access and steady-state sustainment can in principle be achieved via feedback control of the line-averaged electron density by manipulating the pellet fuelling rate and maintaining adequate control of the external heating power.

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.

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

We present the newly discovered control algorithm for the heating power during ignition access to the thermally unstable ignition regime in the FFHR helical reactor. We used the fact that the thermally unstable and stable ignition regimes on Plasma Operating Contour map (POPCON) have an inverse relationship with respect to the density and temperature. As the density limit restricts the operating point to the thermally stable regime, the postulated temperature limit restricts the operating point to the thermally unstable regime. As a result, the heating power for feedback control is expressed by the measured temperature and power balance terms such as alpha heating and bremsstrahlung loss powers. The external heating power thus obtained is applied during the low-density NBI penetration phase, and then switched off automatically after over-passing the saddle point by alpha-heating power, reaching high-density ignition regime. This control algorithm is robust to various perturbations of the confinement factor, impurity content and so on. Feedback control algorithms thus established for both fueling and heating makes the thermally unstable operation more realistic in FFHR.

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.

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.

Based on high-density and high-temperature plasma experiments in the large helical device (LHD), conceptual design studies of the LHD-type helical DEMO reactor FFHR-d1 have been conducted by integrating wide-ranged R&D activities on core plasmas and reactor technologies through cooperative researches under the fusion engineering research project, which has been launched newly in NIFS. Current activities for the FFHR-d1 in this project are presented on design window analyses with designs on core plasma, neutronics for liquid blankets, continuous helical magnets, pellet fueling, tritium systems and plasma heating devices.

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) (Bt = 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 (Bt = 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.

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.

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.

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

Three-dimensional resistive MHD simulation was carried out in order to develop a method to fuel the plasma core by merging two Spherical Torus (ST) plasmas. A relatively small ST plasma was translated into the axial direction by external magnetic field control and made to collide with the larger ST. In this calculation, ballooning instability occurred in both plasmas during translation, while the magnetic surface was observed to collapse from the Poincare plot of the magnetic lines. It was observed that the magnetic lines in the core regions of the two STs were connected to each other and that the merging was completed.

Progress of design studies on a heliotron reactor is reported. Recent experimental achievements in large helical device (LHD) are building up the physical basement. Integrated designs on LHD-type energy reactors have improved the base design of force free helical reactor and revealed economical feasibility. Large helical device is exploring a new horizon of net current free helical plasmas. Inherent characteristics demonstrate reactor relevant performances individually; steady state (1 h), high beta (5%), high density (1 × 10 21 m −3 ), etc. In particular, discovery of internal diffusion barrier provides a novel attractive scenario of super high density operation. It greatly facilitates physical requirements as well as mitigates engineering demands. New methods have been proposed to access and operate ignited plasmas; a long rise-up over 300 s to reduce the heating power to 30 MW and a new proportional-integration-derivative (PID) control of the fueling to handle the thermally unstable plasma at high-density operations.

A new scenario to control transport of 'cold' ?-particles in a helical fusion device by changing poloidal field coil current during plasma discharge is proposed here. A way to enhance radial transport of ?-particles in the intermediate energy range is considered which relies on specific features of helical magnetic field. To remove helically trapped, cold ?-particles, a ?-induced change in the B/B0 modulation along the particle trajectory is employed. To remove passing particles resonant effects connected with the poloidal field coil currents are used.

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

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 β ST plasmas.

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 ? 1017?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?Schl?ter 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.

Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 B477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes Γn ≈ 8.8 × 1013 n s−1 cm−2 ('wall loading' WL ≈ 2 MW m−2), over size-scale >105 cm2 and depth-scale >50 cm, delivering >3 accumulated displacement per atom per year ('neutron fluence' >0.3 MW yr−1 m−2) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~3, Ip ~ 12 MA, BT ~ 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere level. A systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering and technology test missions.

A broadly based study of the fusion engineering and plasma science conditions of a Component Test Facility (CTF), using the Spherical Torus or Spherical Tokamak (ST) configuration, have been carried out. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes {gamma}{sub n} > 4.4 x 10{sup 13} n/s/cm{sup 2}, over size scales > 10{sup 5} cm{sup 2} and depth scales > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year. The desired chamber conditions can be provided by a CTF with R{sub 0} 1.2 m, A = 1.5, elongation {approx} 3, I{sub p} {approx} 9 MA, B{sub T} {approx} 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. Relatively robust ST plasma conditions are adequate, which have been shown achievable without active feedback manipulation of the MHD modes. The ST CTF will test the single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A new systems code that combines the key required plasma and engineering science conditions of CTFmore » has been prepared and utilized as part of this study. The results show high potential for a family of lower-cost CTF devices to suit a variety of fusion engineering science test missions.« less

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.

Recent progress(1) in plasma science of the Spherical Tokamak (or Spherical Torus, ST)(2) has indicated relatively robust plasma conditions in a broad number of topical area including strong shaping, stability limits, energy confinement, self-driven current, and sustainment. This progress has enabled an extensive update of the plasma science and fusion engineering conditions of a Component Test Facility (CTF)(3), which is potentially a necessary step in the development of practical fusion energy. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes n > 4.4 1013 n/s/cm2, over sizescale > 105 cm2 and depth-scale > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year(4). Such chamber conditions are calculated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~ 3, Ip ~ 9 MA, BT ~ 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. The ST CTF will test the life time of single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A newmore » systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering science test missions.« less

Pellet injection algorithm has been proposed for engineering developments in the force free helical reactor (FFHR) based on the zero-dimensional equations. In general, the plasma parameters, such as the density, the temperature and hence the fusion power, may oscillate by repetitive pellet injections during the ignition access and steady state phase. Pellet injection with a higher frequency cannot solely remove oscillation of the fusion power because it is injected irregularly. Gas puffing with the slow time response using the same signal of the fusion power can compensate these oscillations to some level because pellets are injected cyclically. When two pellet systems work alternatively out of phase together with the slow gas puffing, the fusion power oscillation is smallest. In addition, even when the fuel particle confinement time becomes long due to wall saturation, the fusion power and density can be controlled during the fusion burn phase as long as the helium ash can be removed.

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.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7?MW) and high harmonic fast wave (HHFW) heating (up to 6?MW), toroidal fields up to 6?kG, plasma currents up to 1.5?MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10?MW?m?2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400?kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

An overview of steady state tokamak studies in TRIAM-1M (R0 = 0.8 m, a × b = 0.12 m × 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) × 1019 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, ftrans, for the ECD transition is determined as a function of PCD. At ~70 kW no transition occurs for an ftrans value of ~0.017 Hz, meaning almost zero transition probability. With increasing PCD > Pth, ftrans 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 and low power (

The National Spherical Torus Experiment (NSTX) at Princeton University, Princeton, NJ, is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 7 MW of neutral beam injection (NBI) at an injection energy of 100 keV and up to 6 MW of high harmonic fast wave (HHFW) at 30 MHz. NSTX has achieved /spl beta//sub T/ of 32%. A variety of MHD phenomena have been observed to limit /spl beta/. NSTX has now begun addressing /spl tau//sub E/ scaling, /spl beta/ limits, and current drive issues. During the NBI heating experiments, a broad T/sub i/ profile with T/sub i/ up to 2 keV, T/sub i/>T/sub e/ and a large toroidal rotation were observed. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, /spl tau//sub E/ is up to two times the ITER97L L-Mode scaling and exceeds the ITER98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of /spl tau//sub E/ after the transition in some conditions. During HHFW heating, T/sub e/>T/sub i/ and T/sub e/ up to 3.5 keV were observed. Current drive has been studied using both coaxial helicity injection with up to 390 kA of toroidal current and HHFW. HHFW has produced H-modes with significant bootstrap current fraction at low I/sub p/, high q, and high /spl beta//sub p/.

To find a solution of the plasma current rampup problem in a low aspect ratio spherical tokamak (ST) reactor, the effect of the outer vertical field coils on plasma current rampup is studied with noninductively driven current and bootstrap current but without the OH transformer during the fusion power rampup phase. As a lower elongation of [kappa] = 2 does not allow a large poloidal beta, a low density discharge with high heating/current drive power is necessary to increase the noninductive plasma current up to 30 MA, and then the vertical field can ramp the plasma current up to 48 MA just before reaching the steady state fusion burn phase. A higher elongation of [kappa] 3 can reach a higher {beta}{sub p} value, in which case the plasma current of 48 MA can be achieved by the vertical field without powerful heating/current drive. As the level of the vertical field current drive depends on the plasma energy, which is determined by the confinement time, a lower confinement factor can only produce a lower plasma current. The current rampup time can be chosen arbitrarily, shorter or longer, facilitating a flexible ST reactor operation.

Plasma current ramp-up assisted by outer vertical field coils with non-inductively driven current is studied during the ignition access phase in a high aspect ratio tokamak such as ITER-FEAT class tokamak reactor. In a tokamak without the Ohmic transformer, the pl1asma current ramp-up to 8 MA could be achieved by the vertical field with the heating/current drive power of 100 MW. The current ramp-up time can be chosen arbitrarily, as shorter than that expected in the usual, non-inductive current drive operation analyses, and it depends on the fusion power ramp-up time. This makes the reactor operation flexible. Experimental evidence showing the current drive by a vertical field in JT-60U has also been presented with plasma circuit theory.

Fully non-inductive plasma maintenance was achieved by a microwave of 8.2 GHz and 40 kW for more than 1 h 55 min with a well-controlled plasma-facing wall (PFW) temperature of 393 K, using a hot wall in the middle-sized spherical tokamak QUEST, until the discharge was finally terminated by the uncontrollability of the density. The PFW was composed of atmospheric plasma-sprayed tungsten and stainless steel. The hot wall plays an essential role in reducing the amount of wall-stored hydrogen and facilitates hydrogen recycling. The behaviour of fuel hydrogen in the PFW was investigated by monitoring the injection and evacuation of hydrogen into and from the plasma-producing vessel. A fuel particle balance equation based on the presence of a hydrogen transport barrier between the deposited layer and the substrate was applied to the long-duration discharges. It was found that the model could readily predict the observed behaviour in which a higher wall temperature likely gives rise to faster wall saturation.

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 =(dR/dBv)/(dBv/dR) , 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.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

We compare two cases with and without subtraction of dW/dt in the net heating power after the high confinement mode transition (H-transition) and study the effects of dW/dt on zero-dimensional ignition analyses in various ignition phases in a deuterium–tririum (D–T) tokamak reactor including International Thermonuclear Experimental Reactor (ITER). Subtraction of dW/dt from the net heating power after the H-transition as done in H-mode power threshold analyses results in a wrong control action in feedback control during the confinement factor increment and abnormal behaviors in operation phases such as fusion power ramp-up, steady state, and shutdown phases in a reactor. As subtraction of dW/dt in the L-mode phase is to describe the delay time effect of the heating power to the plasma edge, we conclude that dW/dt should not be subtracted from the net heating power after the H-transition in the zero-dimensional approach. This is because that the H-mode is determined by the net heating power to the plasma edge both before and after the H-transition, but not by the plasma conduction loss power out of the plasma edge.

Ignition characteristics are studied using zero-dimensional equations and are compared for both the ITER89P and ITER93HP scalings with the H-mode power threshold in the International Thermonuclear Experimental Reactor-intermediate design (ITER-ID). With the help of the operation path method on the -T plane, the ignition behaviour for both scalings is compared. The comparison study illustrates that the confinement factor of 2.6-2.9 over the ITER89P scaling is necessary for obtaining the equivalent performance to the ITER93HP scaling with 85% degradation due to an ELMy discharge. If the confinement degradation effect of due to the radiative divertor operation is taken into account, the confinement factor should be increased by up to . This result implies that the `advanced' tokamak operation regime may be required even for MA nominal operation from the outset. Modifications of the plasma shape, the central solenoid and divertor would be necessary to increase the confinement time for reducing the ignition risk. The ignition behaviour for both scalings are entirely different, where the ITER93HP scaling is more favourable for the safer ignited operation. Thus, in this sense, ITER89P scaling is highly recommended for designing the ITER.

The temporal evolutions of the operating point during the ignition access and ignited operation phases are analysed on the basis of zero dimensional (O-D) equations in order to clarify the requirements for safe control of ignited operation and for the development of diagnostic systems in ITER. A stable and safe method of reaching the ignited operating point is identified as the `higher temperature access' method, being compatible with the H mode power threshold constraints. It is found that the ignition boundary can be experimentally determined by a `thermonuclear oscillation' of the operating point without knowing the power balance equation. On the other hand, the ignition boundary determined by the power balance equation has a larger error bar depending on the accuracy of the diagnostic system. The plasma waveform response to sudden changes in the various plasma parameters during ignited operation is also calculated, and fusion power regulation is demonstrated by feedback control of the fuelling and auxiliary heating power

Complete one cycle and 1.5 cycle ac operations are performed in the STOR-M tokamak with the plasma current of ∼20 kA using newly developed feedback control and Ohmic heating circuits. Bias voltage adjustment is installed in the plasma position circuit to optimize the plasma position in the second negative plasma current phase for multicycle ac operation. The key to successful, reproducible multicycle ac tokamak operations on STOR-M is to control both the total vertical field by the feedback control system and the plasma position by application of the bias voltage.

Formation of a high performance plasma with βp = 3.6, βN = 1.6, HH98y2 = 1.6 without the use of the Ohmic heating solenoid was demonstrated for the first time. Plasma start-up and plasma current ramp-up were accomplished by vertical field and shaping coils, combined with heating and current drive by EC, LH and NBI. The q profile is deeply reversed with qmin = 5.6 at r/a = 0.7, and the bootstrap current fraction was at least 90%. This result opens up the possibility of eliminating the Ohmic heating solenoid, which has a great impact on improving tokamak and ST reactor designs.

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).

The current rampdown phase followed by non-disruptive current termination is studied for future alternating current (AC) operations in the STOR-M tokamak. The plasma current is smoothly reduced to zero without disruptions with the rampdown rate of around 1.5 kA ms-1. The plasma density in the current rampdown phase is controlled by gas puffing. The plasma current decreases, maintaining the relationship 1/qa approximately 0.01 nR/Bt (10-18 m2T). At current termination, Ip=0, a residual plasma density (0.58-2.9)*1018 m-3 is observed with a corresponding Murakami parameter nR/Bt=(0.41-2.1)*1018 m-2 T-1. After current termination, the low-density plasma remains for 5 to 15 ms, which would allow AC operation in the STOR-M tokamak if the radiation barrier does not affect the subsequent current evolution.

In this paper, the method of the operation path on the generalized ignition contour map ({bar P}{sub ht} {tau}{sup 2}{sub E}, n{tau}{sub E}, T) is used to treat the ignition accessibility of a deuterium-tritium tokamak reactor for different types of confinement scaling including offset linear scaling. It is shown that the operation paths and ignition boundary on the {bar P}{sub ht}{tau}{sup 2}{sub E}-T plane as well as the n{tau}{sub E}-T plane provide a useful ignition criterion for various types of scaling. The International Thermonuclear Experiment Reactor (ITER) tokamaks with low and high toroidal fields are used as examples.

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.

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