The MIT PSFC and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX) [1]-a tokamak specifically designed to address critical needs in the world fusion research program on the pathway to DT fusion devices: 1. Demonstrate robust divertor power handling solutions at reactor-level boundary plasma parameters (heat fluxes, plasma pressures and PMI flux densities), which scale to long-pulse operation 2. Demonstrate nearly complete suppression of divertor material erosion, sufficient to sustain divertor lifetime for ~5x10 7 s of plasma exposure at reactor-level parameters 3. Achieve the above two goals while demonstrating a level of core and pedestal plasma performance that projects favorably to a fusion power plant and in physics regimes that are prototypical 4. Demonstrate efficient radio frequency current drive and heating techniques that solve plasma-material interaction challenges, scale to long-pulse operation and project to effective current profile control 5. Determine high-temperature PMI response of reactor-relevant plasma-facing material candidates, such as tungsten and liquid metals, in an integrated tokamak environment, assessing issues of material erosion, damage, material migration and fuel retention at reactor-level performance parameters. ADX is a high field (≥ 6.5 tesla, 1.5 MA), high power density facility (P/S ~ 1.5 MW/m 2) specifically designed to test innovative divertor ideas at reactor-level plasma/atomic physics parameters-divertor target plate conditions (e.g., T t < ~5eV, n t > ~10 21 m-3 [2]), boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region-while simultaneously producing high performance core plasma conditions prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fueling from external heating and current drive systems. Equally important, the experimental platform is specifically designed to test innovative concepts for lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side-the latter being a location where energetic plasma-material interactions can be controlled and favorable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination-advanced divertors, advanced RF actuators, reactorprototypical core plasma conditions-will enable ADX to explore integrated solutions compatible with attaining enhanced core confinement physics, such as made possible by reversed central shear and flow drive, using only the types of external drive systems that are considered viable for a fusion power plant. Critical need-solution for heat exhaust: As stated in 2013 EFDA report [3]: "A reliable solution to the problem of heat exhaust is probably the main challenge towards the realisation of magnetic confinement fusion...

No abstract

A simple technique for reducing the re-cycling rate and impurity concentration in a tokamak plasma is described. An active metal coating, in this case titanium, evaporated onto the surface of the vacuum vessel, provides a trap for neutral hydrogen and impurity atoms which would otherwise freely penetrate the plasma. With this treatment, the plasma density decays with time after the ionization of the initial filling gas is completed, in contrast to typical standard discharges which have a rising density throughout the entire period of the discharge, indicating a large gas influx. These discharges are observed to have resistances close to that of a pure hydrogen plasma, Zeff ≃ 1.0. There is a corresponding reduction in the intensity of highly ionized spectral lines of oxygen and iron as evidence of reduced impurity concentrations. The value of the effective ion charge, Zeff, can be varied, by pulsing controlled amounts of impurity gases into the hydrogen plasma.

Good confinement of alpha particles in a large magnetic fusion device is a precondition for building a magnetic fusion reactor. The direct measurement of alpha particle losses is of particular interest. Appropriate diagnostics are now being prepared for the Joint European Torus tokamak: a scintillator probe and a set of Faraday cups. Both systems are capable of measuring charged fusion products and ion cyclotron resonance heating tail ions. The design of the lost alpha particle scintillator probe is in the scope of this article. It will allow the detection of particles with a gyroradius between 20 and 140 mm (15% resolution) and a pitch angle between 30° and 86° (5% resolution). As scintillating material P56 will be used. The light emitted by the scintillator caused by charged particles that pass the collimator and hit the scintillator will be detected via a set of optical lenses and a coherent image fiber bundle with a charge coupled device camera and a photomultiplier array. In the following the present design of the scintillator probe with emphasis on the performance of the system, structural resistance against plasma disruptions, and the requirements on the heat protection against plasma and neutral beam induced thermal loads will be described.

Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. The experimental ECCD efficiency increases as the deposition location is moved towards the inboard midplane or towards smaller minor radius for both co and counter injection; the ECCD efficiency also increases with increasing electron density and/or temperature. The experimental ECCD is compared to both the linear theory (TORAY-GA) as well as a quasilinear Fokker-Planck model (CQL3D) and is found to be in better agreement with the more complete Fokker-Planck calculation, especially when the rf power density and/or loop voltage exceed criterion for substantial nonlinear modification of the electron distribution function. The width of the measured ECCD profile is consistent with the theoretically expected width in the absence of radial transport for the current carrying electrons.

No abstract

No abstract

Operation of the S-1 device in a high-current-density (y/AZe> 2x lO"''* A m) regime has created high-electron-temperature spheromaks (50 eV