Understanding and control of electron thermal transport is a critical point of research in magnetic fusion experiments. Previous experiments have shown that operation with reversed magnetic shear (RMS) can suppress electron thermal transport, resulting in the generation of internal transport barriers (ITBs), with the location of the ITB correlated with the location of minimum magnetic shear. The recent upgrades to NSTX-increased magnetic field up to 1 T, increased plasma current up to 2 MA, 2nd neutral beam-present an increased operating space in which to explore electron thermal transport in RMS plasmas. Utilizing TRANSP, we have developed operating scenarios by which to generate RMS in NSTX-U. The results suggest that RMS in NSTX-U can be generated through fast current ramp and early beam injection into a large plasma volume. This is very similar to the procedure that was followed in both TFTR and NSTX to generate RMS. Sustainment of RMS, disregarding non-( qmin = 1) MHD events, requires maintaining a large plasma volume, and increasing the core Te, either via increased plasma current and/or adding heating power. Using this procedure, RMS was sustained for similar to 1 s, with qmin > 1 for that period.

The multipoint Thomson scattering (MPTS) diagnostic system at the National Spherical Torus Experiment Upgrade (NSTX-U) facility is undergoing an upgrade to operate in real-time and interface with the plasma control system (PCS) for NSTX-U. Previous prototyping efforts have shown that spectral analysis and rapid calculations of electron temperature and density are possible on a real-time Linux machine when using up to a 100-Hz laser pulse repetition rate. A remaining challenge was transferring the real-time data to NSTX-U’s PCS, which utilizes the front panel data port (FPDP) protocol. The original proposed method was to convert the real-time data into analog values, but a new solution was developed to keep the output format digital by using an Ethernet controller with a field-programmable gate array (FPGA). This article focuses on a new input module that has been developed to accept incoming user datagram protocol (UDP) packets sent over Ethernet, convert into FPDP format, and integrate into the existing data stream under NSTX-U’s real-time framework.

We report the first observation of global recycling coefficient R near 0.5 in the Lithium Tokamak eXperiment-β (LTX-β), significantly below the minimum R previously reported in other devices. In a series of experiments with varied Li wall conditioning, estimates of the recycling coefficient have been made using a Lyman-α array and DEGAS2 modeling. A progressive reduction in Lyman-α emission with increased lithium and an increase in edge electron temperature are observed. It is also observed that with increasing Li coating thickness, the effective particle confinement time τp* is reduced and approaches TRANSP calculated energy confinement time (τE), with τp* near τE,TRANSP for the lowest recycling coefficients. Edge temperatures approaching core plasma temperatures, first reported in LTX, can now be directly connected to estimates of the recycling coefficient and qualitatively agree with previous UEDGE simulations. The particle flux to the limiting surfaces appears to be significantly reduced in comparison with fluid scrape-off layer (SOL) models, indicating that a large fraction of the SOL ions are mirror trapped. SOL collisionality drops more than an order of magnitude below the banana regime boundary, indicating the importance of kinetic effects. Full-f 1x2v gyrokinetic simulations of SOL field lines with the GKEYLL code indicate that the fraction of ions trapped along field lines increases as collisionality drops, as a result of increased lithium evaporation.

Experimental data from NSTX and DIII-D discharges with the snowflake (SF) divertor configurations are analyzed toward the development of the X-point radiator (XPR) concept. The XPR divertor regime was recently realized in standard divertor configurations in several tokamaks The SF divertor configuration, with an additional poloidal field null nearby the main X-point, could provide additional benefits for the XPR: a higher flux expansion inside the separatrix and an extended private flux region. This may lead to lower temperatures and higher neutral and electron densities, which are thought to be essential for XPR stability, initiation, and impurity containment. In this work, 4 MW NBI-heated H-mode NSTX discharges and 3–5 MW NBI-heated H-mode DIII-D discharges with SF-minus and SF-plus divertors, with the ion B×∇B drift toward the lower divertor, with and without D2 and CD4 seeding, were analyzed. Many experimental XPR features were found, including good or slightly degraded H-mode confinement, significant ELM size reduction, nearly complete divertor power detachment and a significant divertor radiated power loss. However, evidence of the XPR extending into the confined region was inconclusive in the NSTX tokamak, while in DIII-D, a number of discharges demonstrated a stable MARFE-like structure inside the separatrix over a wide operating space. The present analysis supports the SF divertor as a good candidate for further XPR scenario development in DIII-D and NSTX-U.

Recent experiments in the lithium tokamak experiment- β (LTX- β ) have extended the duration, performance, operating conditions, and diagnosis of the flat-temperature profile, low-recycling regime first observed in LTX. As expected, Li retains hydrogen and suppresses edge neutral cooling, allowing increased edge electron temperature, roughly equal to the core T e . Flat temperature profiles had been obtained transiently in LTX, as the plasma density decayed following the cessation of edge gas puffing. Careful control over the fueling in LTX- β has now been shown to sustain the flat T e profile and hot edge unique to the low-recycling regime for multiple confinement times in high performance discharges with decaying or steady density. With low density, the flat T e profile is also seen to extend into the scrape-off layer. Neutral beam heating is observed in target discharges with relatively flat electron temperature profiles ( T edge ∼ T core /2), though beam heating is stronger in discharges with higher fueling, higher density, and depressed edge T e . Beam heating produces additional peaking of the T e profile, without degradation of the energy confinement time. Neutral beam heating of target discharges with relatively flat electron temperature profiles similarly results in broad beam heated temperature profiles. Energy confinement in LTX- β generally compares favorably to ohmic and H-mode scalings, frequently exceeding them by factors of 2–4. New and improved diagnostics in LTX- β enable better characterization of this unique regime, including measurements of ion temperature and high field side Thomson scattering profiles. As an initial step toward characterizing turbulence with no T e gradient and roughly equal density and pressure gradient, core fluctuation spectra have been measured in peaked T e discharges using far-forward scattering and fluctuation reflectometry.

The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 s, and has available up to 15 MW of neutral beam injection power at different tangency radii and 6 MW of high harmonic fast wave heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle (EP) parameter space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, EP stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

Neutral density profiles are measured on the outboard midplane of the National Spherical Torus Experiment Upgrade (NSTX-U) using a two dimensional camera (ENDD, edge neutral density diagnostic) filtered for deuterium Balmer α (D α ) emission interpreted via simulations using the Monte Carlo neutral transport code DEGAS 2. Deuterium atomic densities n D are calculated by inverting the line-integrated D α brightness and using local measurements of electron density n e and temperature T e to determine atomic rate coefficients. The assumptions used in the derivation of n D from D α emissivity are validated using DEGAS 2 to estimate contributions to emissivity due to electron impact excitation and molecular processes. Experimental measurements and DEGAS 2 simulations are compared over a database of L- and H-mode discharges, showing good agreement in D α emissivity profiles. Residual disagreement between experiment and simulations is suggestive of the possible role of intermittent transport and uncertainties in the molecular data. DEGAS 2 simulations are further used to complement the ENDD diagnostic, extracting neutral (atomic and molecular) densities at locations where the ENDD measurement are not made or where the assumptions used in the ENDD analysis are not valid. Deuterium atomic densities and ionization profiles on the outer midplane are compared for L-mode and H-mode discharges in NSTX-U. One-way coupling of DEGAS 2 to UEDGE multi-fluid simulations constrained by experimental data is used to study edge fueling and neutral penetration in NSTX-U discharges.

Electron-scale turbulence (for ) was measured during the L–H transition in the National Spherical Torus Experiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557) using a coherent microwave scattering system. The measurements were carried out at a radial region adjacent to the edge transport barrier (ETB) (at smaller radius than ETB). The observed L–H transition occurred during current flattop, which facilitated the measurement of electron-scale turbulence. The measured electron-scale turbulence is observed to be quasi-stationary before the L–H transition, and an intermittent phase for electron-scale turbulence is observed after the start of the L–H transition with a gradual decrease in overall turbulence density fluctuation spectral power with intermittent large relative variations (on  ∼0.5–1 ms time scale) in the total spectral power. A turbulence-quiescent phase is observed following the intermittent phase, and a significant reduction in the electron-scale turbulence spectral power is only observed at lower wavenumbers, namely –10, which is also seen in different operational NSTX scenarios due to different stabilization mechanisms. A recovery phase is seen after the quiescent phase, where the electron-scale density fluctuation power starts to gradually increase. Simultaneous ion-scale turbulence measurements at larger radius than the electron-scale turbulence measurement location show similar temporal behavior in ion-scale turbulence as in the measured electron-scale turbulence. These observations demonstrate that the suppression of turbulence during the L–H transition is not just limited to the ETB region. None of the measured electron-scale turbulence and ion-scale turbulence from edge into core is found to be obviously leading in the response to the L–H transition, and the overall turbulence suppression after the start of the L–H transition at different radii seems to start at the same time and is a gradual process happening on a tens-of-ms time scale. The trend of decrease in electron-scale turbulence during the L–H transition is found to be consistent with a decrease in the maximum electron-temperature-gradient linear growth rate from linear gyrokinetic stability analysis. However, the observed intermittency in electron-scale turbulence during the intermittent phase cannot be explained by the linear analysis.

A detailed description of a prototype setup for real-time (RT) Thomson scattering (TS) analysis is presented and implemented in the multi-point Thomson scattering (MPTS) diagnostic system at the National Spherical Torus Experiment Upgrade (NSTX-U). The data acquisition hardware was upgraded with RT capable electronics (RT-analog digital converters and a RT server) that allow for fast digitization of the laser pulse signal of eight radial MPTS channels. In addition, a new TS spectrum analysis software for a rapid calculation of electron temperature (Te) and electron density (ne) was developed. Testing of the RT hardware and data analysis software was successfully completed and benchmarked against the standard, post-shot evaluation. Timing tests were performed showing that the end-to-end processing time was reproducibly below 17 ms for the duration of at least 5 s, meeting a 60 Hz deadline by the laser pulse repetition rate over the length of a NSTX-U discharge. The presented RT framework is designed to be scalable in system size, i.e., incorporation of additional radial channels by solely adding additional RT capable hardware. Furthermore, it is scalable in its operation duration and was continuously running for up to 30 min, making it an attractive solution for machines with long discharges such as advanced, non-inductive tokamaks or stellarators.

Plasma turbulence is considered one of the main mechanisms for driving anomalous thermal transport in magnetic confinement fusion devices. Based on first-principle model, gradient-driven gyrokinetic simulations have often been used to explain turbulence-driven transport in present fusion devices, and in fact, many present predictive codes are based on the assumption that turbulence is gradient-driven. However, using the electrostatic global particle-in-cell gyrokinetic tokamak simulation (GTS) code (Wang et al 2010 Phys. Plasmas 17 072511), we will show that while global gradient-driven gyrokinetic simulations provide decent agreement in ion thermal transport with a set of NBI-heated NSTX (Ono et al 2000 Nucl. Fusion 40 557) H-mode plasmas, they are not able to explain the observed electron thermal transport variation in a set of RF-heated L-mode plasmas, where a factor of 2 decrease in electron heat flux is observed after the cessation of the RF heating. Thus, identifying the regime of validity of the gradient-driven assumption is essential for first-principle gyrokinetic simulation. This understanding will help us to more confidently predict the confinement performance of ITER and future magnetic confinement devices.

We report here on observations of magnetic fluctuations in the ion-cyclotron frequency range on NSTX and NSTX-U. In many respects, the fluctuations appear similar to the ion cyclotron emission (ICE) seen in conventional tokamaks. However, a significant difference between previous observations of ICE and the ICE on NSTX is that the frequency of ICE in conventional tokamaks is typically near the ion cyclotron frequency of the energetic fast ions at the plasma edge. In NSTX and NSTX-U, the magnetic fluctuation frequency corresponds to the ion cyclotron frequency deeper in the plasma, near the location of an internal transport barrier. As on conventional tokamaks, higher harmonics of the deuterium cyclotron frequency, as high as the seventh, are seen with the strongest signal sometimes from higher harmonics. The emission usually appears as an irregular sequence of short bursts typically ≤100 μs in duration although nearly continuous emission for several ms has also been seen under some conditions. Measurements of the emission with a toroidal array of fast probes show that the emission is a long wavelength, spatially coherent mode. The emission frequency does not follow an Alfvénic scaling with density, as seen for compressional Alfvén eigenmodes, but does show a linear scaling with local magnetic field strength. The measured emission shows a compressional polarization consistent with a compressional Alfvén wave. No correlation between the neutron rate and the ICE amplitude is seen. Three-wave coupling between the instability responsible for the ICE and lower frequency modes has also been observed.

We present electron-density-sensitive line ratios for Fe xiii– xvi measured in the spectral wavelength range of 200–440 Å and an electron density range of (1–4) × 1013 cm−3. The results provide a test at the high-density limit of density-sensitive line ratios useful for astrophysical studies. The measurements were performed on the National Spherical Torus Experiment-Upgrade, where electron densities were measured independently by the laser Thomson scattering diagnostic. Spectra were collected with a flat-field grazing-incidence spectrometer, which provided a spectral resolution of up to 0.3 Å, i.e., high resolution across the broad wavelength range. The response of the instrument was relatively calibrated using spectroscopic techniques in order to improve accuracy. The line ratios are compared to other laboratory sources and the latest version of CHIANTI (8.0.2), and an agreement within 30% is found.

The spherical tokamak NSTX has been upgraded to include a second neutral beam line, with three independent beam sources, and to be capable of higher toroidal fields and longer duration plasmas (Ono et al 2015 Nucl. Fusion 55 073007). In this paper we describe some of the initial observations of the affect that the higher field and the modified fast-ion distributions have had on the nature of the global Alfvén eigenmodes (GAE). We also report that the GAE excited through a Doppler-shifted ion cyclotron resonance (DCR) were suppressed in a large number of shots with the injection of a small amount of high pitch (V||/V) fast ions, consistent with the predictions of an analytic theory (Gorelenkov et al 2003 Nucl. Fusion 43 228). We show that the experimental scaling of the GAE frequency and toroidal mode numbers with toroidal field is qualitatively consistent with the predictions of the analytic theory, providing validation for the DCR model. The observed suppression of GAE has also been reproduced in simulations with the hybrid ideal stability code HYM (Belova et al 2017 Phys. Plasmas 24 042505).

Reflectometry measurements of compressional (CAE) and global (GAE) Alfvén eigenmodes are analyzed to obtain the amplitude and spatial structure of the density perturbations associated with the modes. A novel analysis technique developed for this purpose is presented. The analysis also naturally yields the amplitude and spatial structure of the density contour radial displacement, which is found to be 2–4 times larger than the value estimated directly from the reflectometer measurements using the much simpler ‘mirror approximation’. The modes were driven by beam ions in a high power (6 MW) neutral beam heated H-mode discharge (#141398) in the National Spherical Torus Experiment. The results of the analysis are used to assess the contribution of the modes to core energy transport and ion heating. The total displacement amplitude of the modes, which is shown to be larger than previously estimated (Crocker et al 2013 Nucl. Fusion 53 43017), is compared to the predicted threshold (Gorelenkov et al 2010 Nucl. Fusion 50 84012) for the anomalously high heat diffusion inferred from transport modeling in similar NSTX discharges. The results of the analysis also have strong implications for the energy transport via coupling of CAEs to kinetic Alfvén waves seen in simulations with the Hybrid MHD code (Belova et al 2015 Phys. Rev. Lett. 115 15001). Finally, the amplitudes of the observed CAEs fall well below the threshold for causing significant ion heating by stochastic velocity space diffusion (Gates et al 2001 Phys. Rev. Lett. 87 205003).

The mission of the spherical tokamak NSTX-U is to explore the physics that drives core and pedestal transport and stability at high-β and low collisionality, as part of the development of the ST concept towards a compact, low-cost ST-based Pilot Plant. NSTX-U will ultimately operate at up to 2 MA and 1 T with up to 10 MW of Neutral Beam Injection (NBI) power for 5 seconds. NSTX-U will operate in a regime where electromagnetic instabilities are expected to dominate transport, and beam-heated NSTX-U plasmas will explore energetic particle (EP) parameter space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. NSTX-U will also develop the physics understanding and control tools to ramp-up and sustain high performance plasmas in a fully-noninductive fashion. NSTX-U began research operations in 2016, but a failure of a divertor magnetic field coil resulted in the suspension of operations and initiation of Recovery activities. During this period, there has been considerable work in the area of analysis, theory and modeling with a goal of understanding the underlying physics to develop predictive models that can be used for highconfidence projections for both ST and higher aspect ratio regimes. The studies have addressed issues in thermal plasma transport, EP-driven instabilities at ion-cyclotron frequencies and below, macrostability, and edge and divertor physics. IAEA-FEC-OV/5-5Ra [Right hand page running head is the paper number in Times New Roman 8 point bold capitals, centred] [Right hand page running head is the paper number in Times New Roman 8 point bold capitals, centred]

Alfvén cascade (cid:1) AC (cid:2) modes are observed in the National Spherical Torus Experiment (cid:3) M. Ono et al. , Nucl. Fusion 40 , 557 (cid:1) 2000 (cid:2)(cid:4) reversed shear plasmas over a wide range (cid:1) up to (cid:5) 25 % on axis, or (cid:5) 11 % at minimum q (cid:2) of (cid:1) (cid:1) ratio of kinetic pressure to magnetic pressure (cid:2) . At low (cid:1) , the AC mode spectrum shows characteristics similar to conventional tokamaks. At higher (cid:1) , distinct (cid:1) and (cid:1) (cid:1) effects are observed in the spectrum, including a significant reduction in the relative size of the frequency sweep and a toroidal mode number dependence in the minimum mode frequency. AC mode structure is obtained using reflectometry. Fast-ion loss associated with AC mode activity is observed. AC mode polarization at the plasma edge is consistent with expectation. Magnetohydrodynamic (cid:1) MHD (cid:2) spectroscopy is shown to be usable to determine q min at both low (cid:1) and high (cid:1) . Observed AC mode structure and frequency are found

A sequence of H-mode discharges with increasing levels of pre-discharge lithium evaporation (‘dose’) was conducted in high triangularity and elongation boundary shape in NSTX. Energy confinement increased, and recycling decreased with increasing lithium dose, similar to a previous lithium dose scan in medium triangularity and elongation plasmas. Data-constrained SOLPS interpretive modeling quantified the edge transport change: the electron particle diffusivity decreased by 10–30x. The electron thermal diffusivity decreased by 4x just inside the top of the pedestal, but increased by up to 5x very near the separatrix. These results provide a baseline expectation for lithium benefits in NSTX-U, which is optimized for a boundary shape similar to the one in this experiment.