Demonstrating improved confinement of energetic ions is one of the key goals of the Wendelstein 7-X (W7-X) stellarator. In the past campaigns, measuring confined fast ions has proven to be challenging. Future deuterium campaigns would open up the option of using fusion-produced neutrons to indirectly observe confined fast ions. There are two neutron populations: 2.45 MeV neutrons from thermonuclear and beam-target fusion, and 14.1 MeV neutrons from DT reactions between tritium fusion products and bulk deuterium. The 14.1 MeV neutron signal can be measured using a scintillating fiber neutron detector, whereas the overall neutron rate is monitored by common radiation safety detectors, for instance fission chambers. The fusion rates are dependent on the slowing-down distribution of the deuterium and tritium ions, which in turn depend on the magnetic configuration via fast ion orbits. In this work, we investigate the effect of magnetic configuration on neutron production rates in W7-X. The neutral beam injection, beam and triton slowing-down distributions, and the fusion reactivity are simulated with the ASCOT suite of codes. The results indicate that the magnetic configuration has only a small effect on the production of 2.45 MeV neutrons from DD fusion and, particularly, on the 14.1 MeV neutron production rates. Despite triton losses of up to 50 %, the amount of 14.1 MeV neutrons produced might be sufficient for a time-resolved detection using a scintillating fiber detector, although only in high-performance discharges.

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...

New results on the I-mode regime of operation on the Alcator C-Mod tokamak are reported. This ELM-free regime features high energy confinement and a steep temperature pedestal, while particle confinement remains at L-mode levels, giving stationary density and avoiding impurity accumulation. I-mode has now been obtained over nearly all of the magnetic fields and currents possible in this high field tokamak (Ip 0.55–1.7 MA, BT 2.8–8 T) using a configuration with B  ×  ∇B drift away from the X-point. Results at 8 T confirm that the L–I power threshold varies only weakly with BT, and that the power range for I-mode increases with BT; no 8 T discharges transitioned to H-mode. Parameter dependences of energy confinement are investigated. Core transport simulations are giving insight into the observed turbulence reduction, profile stiffness and confinement improvement. Pedestal models explain the observed stability to ELMs, and can simulate the observed weakly coherent mode. Conditions for I–H transitions have complex dependences on density as well as power. I-modes have now been maintained in near-DN configurations, leading to improved divertor power flux sharing. Prospects for I-mode on future fusion devices such as ITER and ARC are encouraging. Further experiments on other tokamaks are needed to improve confidence in extrapolation.

Performance predictions for future fusion devices rely on an accurate model of the pedestal structure. The leading candidate for predictive pedestal structure is EPED, and it is imperative to test the underlying hypothesis to further gain confidence for ITER projections. Here, we present experimental work testing one of the EPED hypothesis, namely the existence of a soft limit set by microinstabilities such as the kinetic ballooning mode (KBM). This work extends recent work on Alctor C-Mod [Diallo, et al., Phys. Rev. Lett., 112, (2014), 115001], to include detailed measurements of the edge fluctuations and comparisons of edge simulation codes and experimental observations.

Assessing the performance of lower hybrid current drive (LHCD) at high density is critical for developing non-inductive current drive systems on future steady-state experiments. Excellent LHCD efficiency has been observed during fully non-inductive operation (η = 2.0 − 2.5 × 10 19 AW −1 m −2 atn e = 0.5 × 10 20 m −3 ) on Alcator C-Mod [I. H. Hutchinson, et al, Physics of Plasmas, 1, 1511] under conditions (n e , magnetic field and topology, LHCD frequency) relevant to ITER [S. Shiraiwa, et al, Nuclear Fusion, 51, 103024 (2011)]. To extend these results to advanced tokamak regimes with higher bootstrap current fractions on C-Mod, it is necessary to increasen e to 1.0 − 1.5 × 10 20 m −3 . However, the number of current-carrying, non-thermal electrons generated by LHCD drops sharply in diverted configurations at densities that are well below the density limit previously observed on limited tokamaks. In these cases, changes in scrape off layer (SOL) ionization and density profiles are observed during LHCD, indicating that significant power is transferred from the LH waves to the SOL. Fokker-Planck simulations of these discharges utilizing ray tracing and full wave propagation codes indicate that LH waves in the high density, multi-pass absorption regime linger in the plasma edge and SOL region, where absorption near or outside the LCFS results in the loss of current drive efficiency. Modeling predicts that non-thermal emission increases with stronger single-pass absorption. Experimental data show that increasing T e in high density LH discharges results in higher non-thermal electron emission, as predicted by the models.

Scattering effects induced by edge density fluctuations on lower hybrid (LH) wave propagation are investigated. The scattering model used here is based on the work of Bonoli and Ott (1982 Phys. Fluids 25 361). It utilizes an electromagnetic wave kinetic equation solved by a Monte Carlo technique. This scattering model has been implemented in GENRAY, a ray-tracing code which explicitly simulates wave propagation, as well as collisionless and collisional damping processes, over the entire plasma discharge, including the scrape-off layer that extends from the separatrix to the vessel wall. A numerical analysis of the LH wave trajectories and the power deposition profile with and without scattering is presented for Alcator C-Mod discharges. Comparisons between the measured hard x-ray emission on Alcator C-Mod and simulations of the data obtained from the synthetic diagnostic included in the GENRAY/CQL3D package are shown, with and without the combination of scattering and collisional damping. Implications of these results on LH current drive are discussed.

Experiments on the Alcator C-Mod tokamak have utilized reactor-relevant magnetic fields to sustain substantially higher pedestal pressure than in other devices and allow close approach to the ITER H-mode baseline target pedestal pressure of 90 kPa. The EPED model, which couples the physics of transport driven by kinetic ballooning modes and MHD instabilities arising from peeling-ballooning modes, predicts the pressure profile at the onset of edge-localized modes (ELMs), and yields to lowest order a critical-βN like behavior for the pedestal: ( for fixed edge q). C-Mod routinely accesses edge plasma pressure in excess of 30 kPa, often by using a high-density () approach to high confinement, taking advantage of a regime known as enhanced D-alpha (EDA) H-mode. In the EDA H-mode, plasma transport regulates both the pedestal profiles and the core impurity content, thus holding the pedestal stationary at just below the peeling-ballooning stability boundary. This stationary ELM-suppressed regime has approached the maximum pedestal predicted by EPED at these densities: 60 kPa. This in turn gives rise to volume-averaged core plasma pressure in excess of 0.2 MPa, a world record value for a magnetic fusion device. Another approach to achieving high pressure utilizes a pedestal limited by current-driven modes at low collisionality, in which pressure increases with density and which allows access to a higher EPED solution, termed ‘super-H’. C-Mod experiments at reduced density () and strong plasma shaping () accessed this regime, producing pedestals with pressures up to 80 kPa (approximately 90% of the ITER target) and temperatures of nearly 2 keV. In a number of these hot H-modes, we observe strong edge instabilities at low toroidal mode number (n  =  1) when pedestal pressure approaches predicted values from EPED, showing that current-driven MHD modes can serve as a limit on the pedestal in a metal-walled tokamak at high pressure and low collisionality.

The weakly coherent mode (WCM) in I-mode has been studied by a six-field two-fluid model based on the Braginskii equations under the BOUT++ framework for the first time. The calculations indicate that a tokamak pedestal exhibiting a WCM is linearly unstable to drift Alfven wave (DAW) instabilities and the resistive ballooning mode. The nonlinear simulation shows promising agreement with the experimental measurements of the WCM. The shape of the density spectral and location of the spectral peak of the dominant toroidal number mode n = 20 agrees with the experimental data from reflectometry. The simulated mode propagates in electron diamagnetic direction is consistent with the results from the magnetic probes in the laboratory frame, a large ratio of particle to heat diffusivity is consistent with the distinctive experimental feature of I-mode, and the value of the simulated χe at the edge is in the range of experimental errors of χeff from the experiment. The prediction of the WCM shows that free energy is mainly provided by the electron pressure gradient, which gives guidance for pursuing future I-mode studies.