We describe a fixture that allows electrochemical devices to be studied under electrical bias in the type of vacuum systems commonly used in surface science. Three spring-loaded probes provide independent contacts for device operation and the characterization in vacuum or under in situ conditions with reactive gases. We document the robustness of the electrical contacts over large temperature changes and their reliability for conventional electrochemical measurements such as impedance spectroscopy. The optical access provided to the device enables the analysis by many techniques, as we demonstrate using x-ray photoelectron spectroscopy to measure local electrical potentials on a solid-oxide electrolyte device operating at high temperature in near-ambient pressure.

Research on the National Spherical Torus Experiment, NSTX, targets physics understanding needed for extrapolation to a steady-state ST Fusion Nuclear Science Facility, pilot plant, or DEMO. The unique ST operational space is leveraged to test physics theories for next-step tokamak operation, including ITER. Present research also examines implications for the coming device upgrade, NSTX-U. An energy confinement time, τ E , scaling unified for varied wall conditions exhibits a strong improvement of B T τ E with decreased electron collisionality, accentuated by lithium (Li) wall conditioning. This result is consistent with nonlinear microtearing simulations that match the experimental electron diffusivity quantitatively and predict reduced electron heat transport at lower collisionality. Beam-emission spectroscopy measurements in the steep gradient region of the pedestal indicate the poloidal correlation length of turbulence of about ten ion gyroradii increases at higher electron density gradient and lower T i gradient, consistent with turbulence caused by trapped electron instabilities. Density fluctuations in the pedestal top region indicate ion-scale microturbulence compatible with ion temperature gradient and/or kinetic ballooning mode instabilities. Plasma characteristics change nearly continuously with increasing Li evaporation and edge localized modes (ELMs) stabilize due to edge density gradient alteration. Global mode stability studies show stabilizing resonant kinetic effects are enhanced at lower collisionality, but in stark contrast have almost no dependence on collisionality when the plasma is off-resonance. Combined resistive wall mode radial and poloidal field sensor feedback was used to control n = 1 perturbations and improve stability. The disruption probability due to unstable resistive wall modes (RWMs) was surprisingly reduced at very high β N /l i > 10 consistent with low frequency magnetohydrodynamic spectroscopy measurements of mode stability. Greater instability seen at intermediate β N is consistent with decreased kinetic RWM stabilization. A model-based RWM state-space controller produced long-pulse discharges exceeding β N = 6.4 and β N /l i = 13. Precursor analysis shows 96.3% of disruptions can be predicted with 10 ms warning and a false positive rate of only 2.8%. Disruption halo currents rotate toroidally and can have significant toroidal asymmetry. of this phenomenon in designing future RF systems. The snowflake divertor configuration enhanced by radiative detachment showed large reductions in both steady-state and ELM heat fluxes (ELMing peak values down from 19 MW m −2 to less than 1.5 MW m −2 ). Toroidal asymmetry of heat deposition was observed during ELMs or by 3D fields. The heating power required for accessing H-mode decreased by 30% as the triangularity was decreased by moving the X-point to larger radius, consistent with calculations of the dependence of E × B shear in the edge region on ion heat flux and X-point radius. Co-axial helicity injection reduced the induct...

Plasma facing components for ITER will be manufactured as macro brush structures or with castellated surfaces. Material samples with gaps of similar geometry as intended for ITER were exposed to different plasma conditions in TEXTOR, DIII-D and ASDEX Upgrade. In all devices a decrease of both carbon and deuterium inventories at the side faces from the gap entrance into the gap with scale-lengths in the mm range is found. The fraction of D retained at the gap surfaces is in the range of 0.4-4% of the incident flux. Main parameters determining the retained D-fraction are the temperature of the respective surfaces and the carbon fraction in the incident flux. Extrapolation of tritium inventory growth rates to ITER dimensions assuming the measured retention fractions at T>200ºC and using a D/T-flux distribution with a carbon fraction of from B2/EIRENE simulations of an ITER H-mode discharge yields 1% ≈ 1 a contribution to the increase of the total in-vessel tritium inventory in the range of 0.5-5g T/discharge.

The mission of the National Spherical Torus Experiment (NSTX) is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high β operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of high harmonic fast-waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance l i ∼ 0.4 with strong shaping (κ ∼ 2.7, δ ∼ 0.8) with β N approaching the with-wall β-limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction f NI ∼ 71%. Instabilities driven by super-Alfvénic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvénic. Linear toroidal Alfvén eigenmode thresholds and appreciable fast ion loss during multi-mode bursts are measured and these results are compared with theory. The impact of n > 1 error fields on stability is an important result for ITER. Resistive wall mode/resonant field amplification feedback combined with n = 3 error field control was used on NSTX to maintain plasma rotation with β above the no-wall limit. Other highlights are results of lithium coating experiments, momentum confinement studies, scrape-off layer width scaling, demonstration of divertor heat load mitigation in strongly shaped plasmas and coupling of coaxial helicity injection plasmas to ohmic heating ramp-up. These results advance the ST towards next step fusion energy devices such as NHTX and ST-CTF.

Combining perspectives from the new science of happiness with discussions regarding ''problematic'' and ''addictive'' play in multiplayer online games, the authors examine how player motivations pattern both positive and negative gaming experiences. Specifically, using ethnographic interviews and a survey, the authors explore the utility of Yee's three-factor motivational framework for explaining the positive or negative quality of experiences in the popular online game World of Warcraft (2004-2012). The authors find that playing to Achieve is strongly associated with distressful play, results that support findings from other studies. By contrast, Social and Immersion play lead more typically to positive gaming experiences, conclusions