A comparison between divertor heat loads in ELMy and High Recycling Steady (HRS) H-modes was carried out by fast measurements of the two dimensional D_α emission, and the divertor electron temperature and ion saturation current on JFT-2M. The heat load for the HRS H-mode was maintained at the same level as the time-averaged heat load for the ELMy H-mode, indicating that the HRS H-mode has the attractive advantage of avoiding the severe ELM heat load. An enhancement of particle transport dominated the heat load for the HRS H-mode, whereas the large ELM heat load was induced by abrupt increases of both heat and particle transport.

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Plasma processing, such as etching and deposition, is an indispensable processing technique in the fabrication of modern semiconductor microelectronic devices. Nowadays, increasingly strict requirements are being imposed on plasma processing technology, as integrated circuit device dimensions continue to be scaled down. The numerical simulation is strongly required for a better understanding of the physics and chemistry underlying the processing, and for a design of plasma reactors and plasma processes requiring less experimental efforts. This paper presents an on-going study of the plasma simulation for plasma reactors and the process simulation for microstructural features on substrates, in view of the goal to a technology computer-aided design (TCAD) of plasma processing. Attention is placed on etching in parallel-plate radio frequency (rf) and inductively coupled plasmas, with emphasis on the physical and chemical model used in the simulation.

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High-beta ST research is motivated by its high normalized current I_p/aB_T0 achievable at a low aspect ratio and its direct access to second stability for ballooning modes. Key issues for the high-beta ST sustainment are profile control for ballooning/ current-driven kink stabilities and positioning of the conductive wall for mode suppression. ST termination by resistive wall modes (RWM) revealed that velocity shear is a useful tool for the better stability.

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In order to explore the high-beta attainability of a spherical tokamak, theoretical and simulation researches on the stability and relaxation phenomena problems were reviewed. By classifying various magnetohydrodynamic activities by the impact on the high beta stability, the problems to be solved were identified.

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Auxiliary heated confinement data are being accumulated from medium-size spherical tokamaks MAST and NSTX with aspect ratio A=R/a≈1.5. It has already been demonstrated that high beta (normalized beta β_N > 5) and good confinement (enhancement over tokamak H-mode confinement scaling H_H98y2 > 1) can be achieved simultaneously. Both L-mode and H-mode plasmas have energy confinement times comparable to or exceeding predictions of scaling laws based on conventional aspect ratio tokamak data. Analyses of kinetic profile data indicate that ion transport is near neoclassical in most cases, but electron transport is highly anomalous and is the dominant power loss channel. Indications of internal transport barrier formation are observed in both ion and electron temperature profiles, contributing to the core pressure increase.

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Studies were made on ripple losses of fusion produced alpha particles in a low-aspect-ratio tokamak reactor (VECTOR) by using an orbit-following Monte-Carlo code. Alpha particles are well confined in VECTOR. In a low-aspect-ratio tokamak, the dependence of ripple losses on the number of toroidal-field (TF) coils N is very weak. Assuming a toroidal peaking factor of 2 for the heat load due to loss particles, about 1.5% and 1.0% of TF ripple at the outer edge of plasma might be allowable for the first wall with and without cooling system, respectively. In both cases, the number of TF-coils can be reduced to about 4.

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The present status of experimental and theoretical efforts and future prospects on current start-up and ramp-up methods for ST reactors to reach fusion burning without a center solenoid are discussed. First, we need to form an initial magnetic surface and next to ramp-up the plasma current to a sufficient level for energetic neutral beam injection (NBI). The second heating and current drive by NBI with the ramp-up of the external vertical field may bring the plasma to fusion ignition. The first stage will be realized by refinement of various methods including radio frequency heating and current drive, coaxial helicity injection, plasma merging, and poloidal flux swing. In the second stage, α heating and the bootstrap current dominate as the plasma gets close to burning condition. This is a totally new regime and we need further experimental and theoretical efforts involving novel approaches to ramp-up such a self-regulated plasma successfully.

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Recent spherical tokamak (ST) experiments exhibit many advantageous results including plasma start-up without center solenoid, higher fraction of non-inductive current, and formation of internal transport-barrier as seen on conventional tokamak. In order to reflect these efforts to the design of a so-called ”current ramp-up scenario” and ”non-inductive steady-state (SS) operation scenario” of low aspect-ratio tokamak, fractions of bootstrap current and neutral-beam-driven current on a low-aspect-reactor, VECTOR-OPT, are estimated. The operation with such a SS scenario is shown to be feasible if the normalized plasma beta is raised larger than 5, which is typical in ST, with a broader pressure profile.

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A low aspect ratio tokamak power reactor with super-conducting toroidal field (TF) coils has been proposed. A center solenoid coil system and an inboard blanket were removed. The key point was how to create an engineering design solution for the TF coil system with a high field and high current density. The coil system with a center post radius of less than 1 m can generate a maximum field of 20 T. This coil system yields a compact reactor concept, with plasma major and minor radii of 3.5 m and 1.5 m, respectively, and a fusion power of 3 GW.

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Characteristics of low-aspect-ratio tokamak equilibrium was studied to give a database for the spherical tokamak reactor design with an emphasis of attainable maximum elongation of divertor configurations. Effect of opening of the poloidal coil arrangement around the mid-plane of the machine was also verified. Controllability of the triangularity without center solenoid was analyzed for aspect ratios of 2, 4 and 6 covering spherical to normal tokamaks. At low aspect ratio, the triangularity is limited by the inflation of the plasma surface on the inboard side.

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Considering the geometrical characteristics of tokamak reactors with low aspect ratio, a basic neutronics strategy was derived to construct an inboard structure mainly for neutron shielding and to produce enough tritium in the outboard blanket. The designs for optimal inboard shield were surveyed and the necessary thickness was estimated to make the neutron flux sufficiently low on the super-conducting magnet. In addition, the outer blanket designs were studied to attain a tritium breeding ratio (TBR) sufficiently large for a self-sustaining fusion reactor on the basis of advanced fusion reactor materials.

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Recent results of edge plasma study in spherical tokamaks (ST) are reviewed. Effect of divertor geometry on allowable steady-state power and edge localized mode (ELM) heat load to target plates were studied. Conceptual case study was made on divertor power handling in a low aspect ratio tokamak reactor. By expanding magnetic flux tube and divertor wetted area, it might be possible to handle heat load without radiative cooling in low triangularity configuration.

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Reduced weight of a compact tokamak reactor with low aspect ratio leads to significant advantages in terms of waste and life cycle. This suggests that such a low aspect ratio reactor holds great potential for inspiring the public acceptance of fusion energy in the future energy market.

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Feasibility of D-³He fusion reactor has been investigated by using 0-dimensional power and particle balance equations and the IPB98(y,2) scaling. A self burning spherical tokamak (ST) D-³He reactor requires the high Ti mode and low impurity contamination. The beta value, the confinement improvement factor (H_H98y2) and the superconducting magnet should be improved from the present level.

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Neutral particle diagnostics are widely-used techniques to measure the energy distribution of energetic ions in a plasma for investigating the behavior of them in the plasma. Measurement principle and measurement techniques using neutral particle analyzer are introduced. In relation to an analysis of measured neutral particle fluxes, modeling of the neutral particle distribution in the plasma and a velocity distribution function of energetic ions are described. Several examples of neutral particle measurements and the analysis aiming at understanding energetic ion transport in the plasma are presented.

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A system code for designing helical fusion reactors have been developed. A radial build of a helical reactor is considered in the code, and three-dimensional effect of each type helical reactors is reflected by including detail analysis result of them in other study. Sensitivity analysis have held by using the code so as to clarify the physics and engineering issues needed to be solved to design high performance helical fusion reactor. Here we tried to design two types of helical reactors; i.e., an ignition machine and a commercial reactor. If we design an ignition machine with a major radius of 8˜10 m, the improvement of the energy confinement factor of H_ISS=2.5˜3 might be required. In the commercial helical reactor a high-density/low-temperature operation regime (e.g., n_e˜6×10²⁰m^-3,T_e=8keV)is feasible by improving confinement (H_ISS=2)and by increasing maximum magnetic field strength (B_max=20T), while it is not accessible in tokamak reactors due to the limitations of the Greenwald density and the current drive power.

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The simplified analytical model of a support structure composed of complicated structures such as multiple flexible plates was proposed for dynamic analysis of the ITER major components of the VV and TF coil. The support structure composed of flexible plates and connection bolts was modeled as a spring model composed of only two spring elements simulating the in-plane and out-of-plane stiffness of the support structure with flexible plates including the effect of connection bolts. The stiffness of both these spring models for the VV and TF coil agree well with that of shell models simulating actual structures such as flexible plates and connection bolts. It is therefore found that the spring model with only these two parameters, in-plane and out-of-plane stiffness, simplifies the complicated support structure with flexible plates for the dynamic analysis of the VV and TF coil. Using the proposed spring model, the dynamic analysis of the VV and TF coil for the ITER were performed to estimate the integrity under the design earthquake at Rokkasho, a candidate for an ITER site. As a result, it is found that the maximum relative displacement of 8.6 mm between the VV and TF coil is much less than the current design gap of 100 mm, so that the integrity of the major components (VV and TF coil) of the ITER tokamak device is ensured for the expected earthquake event.

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