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 Γ_n > 4.4×10¹³ n/s/cm², over size-scale > 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, I_p - 9 MA, B_T - 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 new 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.

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An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high β, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high β Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high β and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.

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During 2003-04 MAST has undergone significant enhancements including new divertors, a new central solenoid, error field correction coils, a new inboard gas injection system, a new NB source and several new and enhanced diagnostic systems. A selection of recent results is presented including observation of electron Bernstein wave heating, demonstration of non-solenoid start-up and the first results from error field correction studies.

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Recent Coaxial Helicity Injection (CHI) studies using the Helicity Injected Torus device (HIT-II) have produced discharges with toroidal plasma currents up to 350 kA. Direct measurements using an internal magnetic probe array show a total poloidal flux in the confined plasma significantly greater than the vacuum injector flux, confirming both the unambiguous presence of a closed-flux core region and the generation of poloidal flux through magnetic relaxation. The key innovation for producing these discharges is a sufficiently high ratio of CHI injector current to toroidal field current, quantified as the dimensionless product λ_INJd greater than 0.3, where d is the effective distance between the electrodes and λ_INJ is the inverse magnetic scale length associated with the CHI injector (defined as μ₀I_INJ/ψ_INJ, where I_INJ is the injector current and ψ_INJ is the poloidal injector flux connecting the electrodes). This critical value of λ_INJ d is understood as a balance between two competing processes: reconnection in the injector generating closed poloidal flux, and resistive decay of the closed flux and current. A value of λ_INJd > 0.3 corresponds to a rate of relaxation greater than the rate of resistive decay, allowing the build-up of poloidal flux and toroidal current.

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The favorable properties of the Spherical Torus (ST) arise from its very small aspect ratio. However, small aspect ratio devices have very restricted space for a substantial central solenoid. Thus methods for initiating the plasma current without relying on induction from a central solenoid are essential for the viability of the ST concept. Coaxial Helicity Injection (CHI) is a promising candidate for solenoid-free plasma startup in a ST. Recent experiments on the HIT-II ST at the University of Washington, have demonstrated the capability of a new method, referred to as transient CHI, to produce a high quality, closed-flux equilibrium that has then been coupled to induction, with a reduced requirement for transformer flux [R. Raman, T. R. Jarboe, B. A. Nelson, et al., Phys. Rev. Lett., 075005-1 (2003)]. An initial test of this method on NSTX has produced about 140 kA of toroidal current. Modifications are now underway to improve capability for transient CHI in NSTX.

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This paper describes the ELMs measurement by a fast camera in NSTX. The images obtained by the fast camera reveals the ELMs behavior near the divertor region, and the X-point movement can be seen clearly at a first time. The X-point moves inner and down during large ELM. On the other hand the X-point moves up and down during tiny ELM. The difference of these ELMs behavior is also briefly discussed.

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Electromagnetic instabilities are often excited by fast super-Alfvénic ions produced by neutral beam injection (NBI) in plasmas of the spherical tokamaks START and MAST (toroidal magnetic confinement devices in which the minor a and major R₀ radii of the torus are comparable, R₀/a = 1.2÷1.8). These instabilities are seen as discrete weakly-damped toroidal and elliptical Alfvén eigenmodes (TAEs and EAEs) with frequencies tracing in time the Alfvén scaling with the equilibrium magnetic field and plasma density, or as energetic particle modes (EPMs) whose frequencies don't start from TAE-frequency and sweep down in time faster than the equilibrium parameters change. In some discharges the beam drives Alfvénic-type modes that start from the TAE frequency and sweep in both up- and down- directions. Such electromagnetic perturbations are interpreted as `hole-clump' long-living nonlinear fluctuations of the fast ion distribution function predicted by Berk-Breizman-Petviashvili [Phys. Lett. A238 (1998) 408]. It is found on both START and MAST that the Alfvén instabilities weaken in their mode amplitude and in the number of unstable modes as the pressure of the thermal plasma increases, in agreement with increased thermal ion Landau damping and the pressure effect on core-localised TAEs.

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The main objective of LATE is to demonstrate the formation of Spherical Tokamak (ST) by electron cyclotron heating (ECH) alone without the center solenoid. In the last year, as a result of the extended microwave power from 10 to 30kW and introducing the vertical plasma position control system, the achievable plasma current has been raised substantially compared with the results reported previously⁽¹⁾. With a two-seconds pulse of 2.45GHz microwave power up to 30kW, a plasma current of 2.0kA have been spontaneously initiated under the steady vertical field, then ramped-up slowly to 7.2kA with a slow ramp-up of the vertical field strength (B_v) for the equilibrium of the plasma current loop. The Bv decay index is temporally-controlled during the discharge to successfully increase the plasma elongation with the aid of the active control of the vertical plasma position. This increase in elongation has lead to an effective increase in the plasma current. Magnetic analysis shows that an ST equilibrium, having the last closed flux surface with an aspect ratio of R₀ / a = 20.5cm/14.8cm 1.4, an elongation of κ 1.6 and q_edge 34, has been finally formed and maintained steady until the end of the microwave pulse.

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Spontaneous formation of spherical tokamak is observed during a microwave discharge at the electron cyclotron resonance (ECR) under a steady vertical magnetic field. In the course of slow plasma current increase, a fast rise of current (usually within several ms) occurs and the magnetic field topology changes drastically from open field type to closed one. After this current jump, a steady plasma current is maintained. The plasma current in the steady stage is proportional to the strength of the vertical field which balances the outward hoop force of the plasma current and maintains the MHD equilibrium. When a 5GHz, 130kW, 60ms microwave power is injected at 85G vertical field, plasma current of 6.8kA is obtained.

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Spherical tokamak program in China was started up from 1999. The Sino United Spherical tokamak (SUNIST) has been assembled in November 2002. Test discharge of SUNIST completed at the end of 2002. We got the plasma with about 50 kA of current in test discharge without flattop on plasma current. After modification of the power supply of vertical field in 2003, we obtained fine equilibrium plasma current on SUNIST with about 2 ms flattop. The SUNIST laboratory has been founded in 2004, consisted of Department of Engineering Physics, Tsinghus University (DEPTS.) and Institute of Physics, Chinese Academy of Sciences (IOPCAS). A series of experiments has been taken on edge plasma parameters, fluctuation and turbulence before and after power supply modification. At the end of 2003, we tried to deposit siliconized film on vacuum vessel. After siliconization, plasma current flattop could extend to the regime where the signal of loop flux had fell down to zero.

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In the edge plasma of the SUNIST spherical tokamak, the toroidal flow velocity v_φ, the poloidal shear layer and the fluctuation-driven particle transport fluxes Γ_r have been simultaneously measured with Mach probe and Langmuir probe arrays, respectively. The results indicate that the v_φ has a radial gradient dv_φ/dr. The change of the dv_φ/dr positively correlates with both the poloidal shear and the suppression of the Γ_r. Such coincidence suggests that the toroidal sheared flow may play an important role in the suppression of the edge fluctuation-driven transport in STs.

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Numerical simulation which is based on the nonlinear resistive magnetohydrodynamic(MHD) model is executed to reveal the dynamics of the edge-localized modes(ELMs) in the spherical tokamak(ST) plasma. Characteristic features of the ELMs, such as the formation of the filaments, the toroidal localization of the mode structures, and the separation of plasmoid from the bulk, have been successfully reproduced. These processes are found to be the consequence of the nonlinear growth of the ballooning mode under the MHD assumptions. The simulation results are compared with the experimental observations in Mega-Amp Spherical Tokamak(MAST), and show good agreement.

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Studies on the loss of fusion produced alpha particles enhanced by toroidal field (TF) ripple in a low-aspect-ratio tokamak reactor (VECTOR) have been made by using an orbit-following Monte-Carlo code. The ripple loss is strongly reduced as the aspect ratio becomes low. Consequently, alpha particles are well confined in VECTOR. In a low-aspect-ratio system, the dependence of the ripple loss on the number of TF coils is very weak, if the edge field ripple is kept constant. Thanks to the good confinement of alphas in a low-aspect-ratio system, the number of TF coils can be reduced to about 6, one half of the original VECTOR, by installing cooling systems near the outer edge of plasma and making allowances for about 30% increase in the bore diameter of TF coils.

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

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Steady burning criteria for a D-³He fusion reactor have been investigated by using the 0-D power and particle balance equations and the IPB98(y,2) scaling law. A spherical tokamak (ST) configuration with the internal transport barrier (ITB) and the bootstrap current fraction of 100 % are assumed. As main results, the hot ion mode (T_i/T_e > 1) and the high central beta are essential to realize a D-³He fusion reactor, and the confinement improvement does not reduce the requirement of the hot ion mode. The reactor size, the magnetic field, confinement improvement factor and the heat flux may be attainable if the hot ion mode is realized.

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We studied transmutation of high-level wastes in a spherical-tokamak reactor. We examined spent fuel, recovered fuel, natural UO₂ fuel and reprocessed waste, though the natural UO₂ fuel is not a high-level waste. We added a few percent of ²³⁹Pu to the former three fuels to increase the tritium breeding ratio and saw the change in the effective multiplication factor k_eff, number of fissions, multiplication of fusion energy and conversion ratio of ²³⁹Pu. We found that 2.5 and 5.5 fissions are needed to obtain the tritium breeding ratio of 1 and 2, respectively. High amplification of fusion energy occurs. Breeding of ²³⁹Pu is expected for the natural UO₂ fuel even in a subcritical environment. For the reprocessed waste, ⁹³Zr and ⁹⁹Tc are transmuted efficiently by the (n,γ) reaction and so are ²³⁷Np and ²⁴¹Am by the (n,f) reaction. Fusion neutrons of 100MW will transmute half of those nuclides produced from the generation of 1 GW_(e) year in a large pressurized water reactor. However, tritium fuel may not be self-sustained in the reprocessed-waste burner.

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The high-power reconnection heating of spherical tokamak (ST) has been studied in the TS-3 experiment by use of axial merging of two STs. In this method, the merging/ magnetic reconnection transformed a part of magnetic energy of merging STs into their thermal energy within short reconnection time. Our present low-field merging (0.3-0.5kG, R-0.2m) attained the maximum heating power of 4-10MW and increased the beta-values of STs by factor 2-3 transiently. The ion heating energy was found to increase inversely with the q-value (B_t component) of two STs. The most probable cause for this dependence is fast reconnection speed/ outflow speed due to large anomalous resistivity of current sheet in low-q state. After merging startup, decrease in beta was observed especially in low-q STs, indicating that the final beta value of ST increased with the q-value.

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A fast framing camera measurement has been performed in order to investigate dynamics of current reversal of helicity-driven spherical torus (ST) plasma. Fast camera image shows the helical kink behavior of the open flux around the central conductor during the relaxation transition from ST to Flipped ST plasmas. This is in good agreement with results of three-dimensional MHD numerical simulation previously reported.

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Two-fluid flowing equilibria of a helicity-driven spherical torus (HD-ST) are numerically investigated by using the combination of the finite difference and the boundary element methods. It is shown from the numerical results that electron fluids near the central conductor are tied to an external toroidal field and ion fluids are not. The magnetic configurations change from the high-q HD-ST (q>1) with paramagnetic toroidal field and low-β (volume average β value, <β>≈2 %) through the helicity-driven spheromak and reversed-field pinch to the ultra low-q HD-ST (0<q<1) with diamagnetic toroidal field and high-β (<β>≈18 %) as the vacuum toroidal field at the inner edge regions decreases and reverses sign. The effects of the toroidal ion flow on the magnetic configurations are discussed.

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This paper presents a proposal of a spherical tokamak (ST) with the superconducting center solenoid and toroidal coil system having the total magnetic stored energy comparable to that of the Large Helical Device. The parameters are as follows: R=1.8-2.1 m, B_t0=2-2.4 T, I_p=8-12 MA (I_OH=4-6 MA), T_e0=8-12 keV, n_e0-4x10²⁰ m^-3 and Q(equivalent)=0.3-infinite depending on the transport scaling. This machine is suitable to study the role of the aspect ratio in the physics (cf. the stability condition of the high beta disruption) of the steady state plasma, which is driven by the bootstrap current.

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Equilibria of the steady state ST up to β_p=1.7 (β_N=4.45) are analyzed from the viewpoint of its engineering realization. The central solenoid (CS) shorter than the vertical plasma size is favorable to suppress excessive inflation of the inboard plasma surface. Staggered central solenoids are required to attain the high elongation and the high triangularity.

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