Several improvements to the MAST plant and diagnostics have facilitated new studies advancing the physics basis for ITER and DEMO, as well as for future spherical tokamaks. Using the increased heating capabilities P NBI ≤ 3.8 MW H-mode at I p = 1.2 MA was accessed showing that the energy confinement on MAST scales more weakly with I p and more strongly with B t than in the ITER IPB98(y,2) scaling. Measurements of the fuel retention of shallow pellets extrapolate to an ITER particle throughput of 70% of its original design value. The anomalous momentum diffusion, χ φ , is linked to the ion diffusion, χ i , with a Prandtl number close to P φ ≈ χ φ /χ i ≈ 1, although χ i approaches neoclassical values. New high spatially resolved measurements of the edge radial electric field, E r , show that the position of steepest gradients in electron pressure and E r are coincident, but their magnitudes are not linked. The T e pedestal width on MAST scales with the β pol rather than ρ pol . The ELM frequency for type-IV ELMs, new in MAST, was almost doubled using n = 2 resonant magnetic perturbations from a set of 4 external coils (n = 1, 2). A new internal 12 coil set (n ≤ 3) has been commissioned. The filaments in the inter-ELM and L-mode phase are different from ELM filaments, and the characteristics in L-mode agree well with turbulence calculations. A variety of fast particle driven instabilities were studied from 10 kHz saturated fishbone like activity up to 3.8 MHz compressional Alfvén eigenmodes (CAE). The damping rate of ellipticityinduced AE was measured to be 4% using the new internal coils as antennae. Fast particle instabilities also affect the off-axis NBI current drive and lead to fast ion diffusion of the order of 0.5 m 2 /s and reduce the driven current fraction from 40% to 30%. EBW current drive start-up is demonstrated for the first time in a spherical tokamak generating plasma currents up to 55 kA. Many of these studies contributed to the physics basis of a planned upgrade to MAST. Introduction: MAST [1]is one of the two leading tight aspect ratio (A = ε −1 = R/a = 0.85 m/0.65 m ∼ 1.3, I p ≤ 1.5 MA) tokamaks in the world. The hot T ≤ 3 keV, dense n e = (0.1 − 1) × 10 20 m −3 and highly shaped (δ ≤ 0.5, 1.6 ≤ κ ≤ 2.5) plasmas are accessed at moderate toroidal field B t (R = 0.7 m) ≤ 0.62 T and show many similarities to conventional aspect ratio tokamaks. Detailed physics studies using the extensive array of state of the art diagnostics and access to different physics regimes help to consolidate the physics basis for ITER and DEMO [2,3], and explore the viability of future devices based on the spherical tokamak (ST) concept such as a component test facility (CTF) [4] or an advanced power plant [5]. The challenge for today's experiments is to find an integrated scenario that extrapolates to these future devices, in particular to develop plasmas with reduced power load on plasma facing components, notably from edge localised modes (ELM), but high confinement facilitated by internal or edge transport ba...

predicts the observed scaling of the low frequency limit for CAE.

A high harmonic fast wave (HHFW) antenna array, designed to provide up to 6 MW of power at 30MHz for heating and current drive applications, has been operated on the NSTX experiment at Princeton Plasma Physics Laboratory (PPPL). The full array consists of twelve evenly spaced, identical current strap modules connected in pairs. Each pair is connected as a half-wave resonant loop and is intended to be driven by one transmitter, allowing rapid phase shift between transmitters. A decoupling network compensates for the mutual inductive coupling between adjacent current straps, effectively isolating the six transmitters from one another. Initial rf operation between November 1999 and January 2000 used eight straps to form four loops, which were driven by two transmitters. Two adjacent loops were connected with a λ/2 coax section to be driven out of phase by a single transmitter. Up to 2 MW of power was delivered during this stage of operation; inter-loop phasings of 0−π−π−0 and 0−π−0−π were investigated. Models of the power distribution system indicate the nominal plasma loading was about 5 Ω/m, close to the design value of 6 Ω/m. The HHFW system has now been reconfigured for 12-strap, 6-transmitter operation with decouplers; low power vacuum and plasma measurements have begun.

DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter-I p steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high-Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L–H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at ∼8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co-I p beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate β N in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation.

No abstract

Abstract.Saturated internal kink modes have been observed in many of the highest toroidal β discharges of the National Spherical Torus Experiment (NSTX). These modes often cause rotation flattening in the plasma core, can degrade energy confinement, and in some cases contribute to the complete loss of plasma angular momentum and stored energy. Characteristics of the modes are measured using soft X-ray, kinetic profile, and magnetic diagnostics. Toroidal flows approaching Alfvénic speeds, island pressure peaking, and enhanced viscous and diamagnetic effects associated with high-β may contribute to mode non-linear stabilization. These saturation mechanisms are investigated for NSTX parameters and compared to experimental data.

Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction are systematically determined as a function of plasma aspect ratio. For plasmas with and without wall stabilization of external kink modes, the computed limits are well described by distinct and nearly invariant values of a normalized beta parameter utilizing the total magnetic field energy density inside the plasma. Stability limit data from the low aspect ratio National Spherical Torus Experiment is compared to these theoretical limits and indicates that ideal non-rotating plasma no-wall beta limits have been exceeded in regimes with sufficiently high cylindrical safety factor. These results could impact the choice of aspect ratio in future fusion power plants. Introduction -The superconducting advanced tokamak [1, 2] is presently the leading candidate for producing an efficient magnetic fusion reactor. Alternative concepts such as the compact stellarator [3,4] and spherical torus [5][6][7] are also actively being pursued as possible improvements to the advanced tokamak. The advanced tokamak (AT) and spherical torus (ST) reactor concepts have several features in common. In particular, both rely on the neoclassical bootstrap current [8] to sustain nearly all of the plasma current and on stabilization of pressure-driven external kink modes to achieve sufficiently high beta (ratio of plasma kinetic pressure to magnetic pressure) to produce power efficiently. The AT and ST reactor concepts have been independently optimized for various physics and engineering constraints and arrive at notably different plasma aspect ratio and beta. This difference has motivated the present work which seeks to understand how the theoretical ideal magnetohydrodynamic (MHD) stability limits of the AT and ST are linked. More generally, aspect ratio invariants of stability are sought.The first equilibrium regime studied consists of fully self-driven plasmas utilizing a close-fitting conducting wall to stabilize external kink modes. The stability limits of this regime represent the maximum achievable beta for this class of equilibrium at any aspect ratio given the present understanding of the relevant physics. The second regime studied consists of plasmas with a self-driven current fraction of 50% and no conducting wall stabilization of the external kink mode. The stability limits of this regime have largely been experimentally realized in present-day tokamaks [9] but have only recently been realized in relatively new ST experiments. Finally, the no-wall current limit is studied for typical AT and ST configurations. These studies show that the degeneracy

Abstract. Global magnetohydrodynamic stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized beta and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during normal operation and increased the maximum achievable beta. As a result of these improvements, peak beta values have reached (not simultaneously) β t = 35%, β N = 6.4, β N = 4.5, β N /l i = 10, and β P = 1.4. High β P operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 second with sustained periods of β N ≈ 6 above the ideal no-wall limit and near the with-wall limit. Details of the β limit scalings and β-limiting instabilities in various operating regimes are described.