High-temperature superconductors (HTS) promise to revolutionize high-power applications like wind generators, DC power cables, particle accelerators, and fusion energy devices. A practical HTS cable must not degrade under severe mechanical, electrical, and thermal conditions; have simple, low-resistance, and manufacturable electrical joints; high thermal stability; and rapid detection of thermal runaway quench events. We have designed and experimentally qualified a Vacuum Pressure Impregnated, Insulated, Partially transposed, Extruded, and Roll-formed (VIPER) cable that simultaneously satisfies all of these requirements for the first time. VIPER cable critical currents are stable over thousands of mechanical cycles at extreme electromechanical force levels, multiple cryogenic thermal cycles, and dozens of quench-like transient events. Electrical joints between VIPER cables are simple, robust, and demountable. Two independent, integrated fiber-optic quench detectors outperform standard quench detection approaches. VIPER cable represents a key milestone in next-step energy generation and transmission technologies and in the maturity of high-temperature superconductors as a technology.

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...

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only highpower radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing componentsapproaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated a) Paper AR1 1, Bull. Am. Phys. Soc. 58, 21 (2013). b) Invited speaker.