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Open field line currents are intrinsic to DC helicity injection plasma startup and pose a challenge for inferring the plasma equilibrium with standard reconstruction analysis. Local helicity injection (LHI) is a type of DC helicity injection which uses small, modular current sources to drive force-free current along helical field lines to produce tokamak plasmas. MHD modeling and magnetic measurements during LHI indicate the injected current streams remain coherent as helical structures on the outboard edge of a core toroidal plasma that is tokamak-like in a toroidally averaged sense. To extract core plasma equilibrium properties, external magnetic diagnostics corrected for contributions from the injected current streams are fitted by a standard Grad-Shafranov equilibrium code. An iterative approach for estimating and subtracting the stream contributions from the diagnostic signals is described and applied to a model equilibrium database to reduce systematic errors introduced by the streams. Convergence is usually attained with 2 to 4 iterations, with derived equilibrium parameters matching the prescribed axisymmetric core values to within estimated experimental uncertainties. Accurate recovery of core parameters occurs when the ratio of the net toroidal windup current from the streams to the core plasma current is less than 0.2, which is typically satisfied in most experiments.

Solenoid-free tokamak startup via point-source DC helicity injection is demonstrated on the Pegasus Toroidal Experiment using a high current density, low impurity plasma gun mounted near the outboard midplane. A threshold in the vacuum vertical magnetic field strength that allows the injected current filament to relax into a tokamak-like topology is observed. A simple 2-D model of the vacuum magnetic field suggests this threshold is the maximum field strength that allows a toroidally connected field null to form. Discharges with I p ≈ 17 kA are produced using less than 2 kA of injected current and no inductive drive. The tokamak-like discharges exhibit current decay times about five times longer than the injected current decay, expansion of the plasma into the vacuum region and a significant increase in the line-integrated density.

Formation of tokamak-like plasmas via electrostatic helicity injection in the ultra-low aspect ratio Pegasus Toroidal Experiment is reported. Two low-impurity, high-current (1 kA) washer gun current sources have been installed in the lower divertor region. These initially drive current along helical field lines produced by the applied toroidal and vertical fields. At sufficiently low values of externally applied vertical field, the poloidal field generated by the plasma is large enough to cause a poloidal flux reversal. In these cases the plasma relaxes into a tokamak-like configuration. Discharges with Iϕ≈ 30 kA are produced with less than 2 kA of injected current. These discharges exhibit features indicative of tokamak plasmas, including reversal of poloidal flux at the center column, strong vacuum field deformation, increased current decay times, increased core heating, and characteristic MHD modes common to other helicity-injection-driven toroidal devices.

Recent Pegasus experiments are developing solenoid-free startup techniques using point-source magnetic helicity injection. These plasma sources, called “plasma guns”, ionize a stream of gas in a discharge channel, and bias this channel with respect to an external electrode, driving current along the plasma stream, which relaxes into a tokamak-like equilibrium. The relaxed discharges formed by these injectors exhibit high current amplification, which is the ratio of total toroidal current to the gun-driven current. The development and present design of these injectors are described, and time traces from a typical discharge are presented.

Large stable values of normalized current I N are achievable in spherical tori due to the large gradient of the toroidal field across the plasma. This allows access to high Troyon beta limits. Values of I N  > 12 MA/m-T are expected to be stable to ideal MHD modes in the ultra-low-A Pegasus Toroidal Experiment. In previous experiments, the toroidal field utilization (I p /I tf ) was found to be limited roughly to unity by the onset of large-scale low-order tearing modes during the current ramp, which limited I N to roughly 6 MA/m-T. Three techniques have now been developed that allow access to higher values of I N , primarily through modification of the current density profile. The first of these techniques employs a fast rampdown of the low-inductance toroidal field coil to rapidly decrease I tf . The other techniques employ gas-fueled washer stack guns as plasma sources. The second technique uses these guns to provide preionization at low toroidal field, while the third relies on the guns as helicity sources to form ST plasmas non-inductively. Using these techniques, values of I p /I tf  > 2 have been obtained.

Detailed 2D turbulence measurements from the DIII-D tokamak provide an explanation for how resonant magnetic perturbations (RMPs) raise the L-H power threshold \\(P_\\textrm{LH}\\) [P. Gohil et al., Nucl. Fusion 51, 103020 (2011)] in ITER-relevant, low rotation, ITER-similar-shape plasmas with favorable ion \\(\\nabla B\\) direction. RMPs simultaneously raise the turbulence decorrelation rate \\(\\Delta \\omega_D\\) and reduce the flow shear rate \\(\\omega_\\textrm{shear}\\) in the stationary L-mode state preceding the L-H transition, thereby disrupting the turbulence shear suppression mechanism. RMPs also reduce the Reynolds stress drive for poloidal flow, contributing to the reduction of \\(\\omega_\\textrm{shear}\\) On the ~100 {\\mu}s timescale of the L-H transition, RMPs reduce Reynolds-stress-driven energy transfer from turbulence to flows by an order of magnitude, challenging the energy depletion theory for the L-H trigger mechanism. In contrast, non-resonant magnetic perturbations, which do not significantly affect \\(P_\\textrm{LH}\\), do not affect \\(\\Delta \\omega_D\\) and only slightly reduce \\(\\omega_\\textrm{shear}\\) and Reynolds-stress-driven energy transfer.

Initial experiments in the Pegasus ST exhibited an operational limit of Ip ∼ Itf at A ∼ 1.15. With new power supplies, discharge control has greatly increased. Equilibria and stability modeling of the high Ip/Itf regime has produced stable equilibria with Ip/Itf up to 2 while values of Ip/Itf approaching 3 show a potential upper bound on stability. A new operating regime has been accessed in recent experiments with the addition of electrostatic plasma sources. Using the sources as a pre-ionization technique, Ip/Itf values of 1.5 have been accessed at very low toroidal fields. Using the sources as a means for non-inductive startup, values of Ip/Itf up to 2.3 are attained. The MHD activity in these discharges is characteristically different than ohmic discharges, suggesting the new current sources are modifying the current profile to allow stable discharge evolution at minimal toroidal field.

The Department of Energy (DOE) Office of Energy Research chartered through the Fusion Energy Sciences Advisory Committee (FESAC) a panel to “address the topic of U.S. participation in an ITER construction phase, assuming the ITER Parties decide to proceed with construction”. Given that there is expected to be a transition period of 3 to 5 years between the conclusion of the Engineering Design Activities (EDA) and the possible construction start, the DOE Office of Energy Research expanded the charge to “include the U.S. role in an interim period between the EDA and construction”.This panel has heard presentations and received input from a wide cross-section of parties with an interest in the fusion program. The panel concluded it could best fulfill its responsibility under this charge by considering the fusion energy science and technology portion of the U.S. program in its entirely. Accordingly, the panel is making some recommendations for optimum use of the transition period considering the goals of the fusion program and budget pressures.