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In the Force Free Helical Reactor FFHR-d1 (R∼15.7 m, a∼2.5 m, Bo ∼4.5 T, 〈β〉∼5 % and the fusion power of 3 GW) [1, 2], it is demonstrated that the thermally unstable operation is better to achieve the higher density and lower temperature plasma (n(0)∼ 9 × 1020 m−3 and T(0)∼7 keV) than the impurity injection method. One drawback of the thermally unstable operation is a possibility of the thermal runaway when fueling systems have failures. However, we have found that the inherently safe function exists owing to the plasma outward shift during the thermal runaway. Thus we continue to pursue this high-density operation scenario for FFHR-d1. Preliminary estimation of the alpha energy loss fraction of ∼3 % is obtained for a broad density profile in the high-density ignition regime.

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.

Based on the various studies on the ignition characteristics, the feedback control algorithm for ignition burn has been developed for a safe operation and diagnostic integration design of ITER. A set of the diagnostic systems for core burn control are selected and a fail-safe control algorithms has been proposed to prevent the unsafe operation caused by the diagnostic failure. Some feature in the reduced cost of ITER Is also presented.

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 (HH98y2) and the superconducting magnet should be improved from the present level.

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.

Electron Bernstein wave heating experiments based on the X-mode to Bernstein mode mode-conversion scenario were performed on the Tokyo Spherical Tokamak -2 (TST-2). Up to 140 kW of microwave power at 8.2 GHz was injected perpendicularly from the low-field side. Evidence of electron heating was observed as increases of the stored energy (by about 15%) and soft X-ray (hν > 1 keV) emission.

Plasma current start-up and ramp-up to 10 kA have been demonstrated in the TST-2 spherical tokamak without the use of the central solenoid. Only the electron cyclotron wave (ECW) and the outer equilibrium field coils are used. The plasma current evolution depends on the poloidal coil arrangement. It is also demonstrated that the plasma current start-up can take place without the field null.

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).

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.

Alpha-particle slowing-down behaviors at low-temperature, high-density operation points in force-free helical reactors (FFHRs) are examined on the basis of a Fokker-Planck (FP) simulation that simultaneously consider the balance among generation, slowing down, and loss from the plasma in parallel with the density dependence of the Alfven speed. An accurate treatment of the boundary velocity region between thermal and non-thermal components is shown to be important in evaluating the alpha particle population that can induce instability. In a typical high-density, low-temperature operation point in an FFHR, this population is reduced.

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.

New control method of the unstable operating point in the helical reactor FFHR makes the ignition study on the high density and low temperature operation possible. Proportional-integral-derivative (PID) control of the fueling with the error of the fusion power of e'DT(Pf) = −(Pfo − Pf) can stabilize the unstable operating point. Here Pfo(t) is the fusion power preset value and Pf(t) is the measured fusion power. Although the large parameter variation would lose its control due to the inherently unstable nature, it is possible to control the ignited operation by pellet injection with the pellet size between 12 mm and 16 mm. Unstable ignited operation is robust against disturbances such as impurity increments by fueling feedback alone. However, if the heating power feedback control is added, robustness to the disturbances is improved, and an operational regime with respect to the integration time and derivative time is expanded.

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

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

A new and simple control method of the unstable operating point in the Force Free Helical Reactor (FFHR) is proposed for the ignited operation with high-density plasma. Proportional-integration-derivative (PID) control of the fueling has been used to obtain the desired fusion power with the fusion power error of e(Pf) = (Pfo − Pf) in the stable operating point. We have discovered that in the unstable regime the error of the fusion power with an opposite sign of e(Pf) = −(Pfo − Pf) can stabilize the unstable operating point. Around the unstable operating point, excess fusion power supplies fueling and then increases the density and decreases the temperature. Less fusion power in the sub-ignited regime reduces the fueling, decreases the density, and increases the temperature. The operating point approaches the final unstable operating point as oscillation is damped away.