National Spherical Torus Experiment Research Articles

Optimization of the snowflake divertor for power and particle exhaust on NSTX–U

Abstract In this paper, simple analytical modeling and numerical simulations performed with the multi-fluid edge transport code UEDGE are used to identify optimal snowflake divertor (SFD) configurations for heat flux mitigation and sufficient cryopumping performance on the National Spherical Torus eXperiment Upgrade (NSTX–U). A model is presented that describes the partitioning of sheath-limited SOL power and particle exhaust in the SFD as a result of diffusive transport to multiple activated strike points. The model is validated against UEDGE predictions and used to analyze a database of 70 SFD-minus equilibria. The optimal location for the entrance to a divertor cryopumping system on NSTX–U is computed for enabling sufficient pumping performance with acceptable power loading in a variety of SFD-minus configurations. UEDGE simulations of one promising equilibrium from the database indicate that a significant redistribution of power to the divertor legs occurs as a result of neutral particle removal near one of the SFD-minus strike points in the outboard scrape-off layer. It is concluded that pump placement at the optimal location is advantageous as the large number of compatible equilibria reduces the precision required of real-time SFD configuration control systems and enables acceptable divertor solutions even if UEDGE-predicted power redistribution slightly reduces the achievable pumping performance.

A scalable real-time framework for Thomson scattering analysis: Application to NSTX-U.

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.

Resolving interactions between ion-cyclotron range of frequencies heating and the scrape-off layer plasma in EAST using divertor probes**This article is dedicated to the memory of Joel C Hosea.

Waves in the ion-cyclotron range of frequencies (ICRF) used for tokamak heating are known to interact strongly with the scrape-off layer (SOL) plasma, potentially limiting ICRF performance. We present the effect of ICRF heating on triple-probe signals in the Experimental Advanced Superconducting Tokamak (EAST) experiment, an advanced superconducting tokamak with two ICRF antennas and 12 MW of ICRF source power in support of high-power and long-pulse operation. Applying time-resolved magnetic mapping to the divertor probe arrays enables detailed sensing of the spatial structure of the SOL modifications. We observe a sharp transition in floating potential from negative to positive over a length scale of a centimeter at a major radius just inboard of the ICRF antenna. The largest changes in floating potential are observed not for probes connecting to the active ICRF antennas but rather to the main limiter or limiters for inactive lower hybrid launchers. Ion saturation current can be greatly reduced for probes mapping just in front of an antenna. Where relevant, comparison with results from the National Spherical Torus Experiment are made. This work will provide insight into sputtering, impurity production and antenna coupling as higher ICRF power levels are attempted on EAST.

TRANSP-based closed-loop simulations of current profile optimal regulation in NSTX-Upgrade

Active control of the toroidal current density profile is critical for the upgraded National Spherical Torus eXperiment device (NSTX-U) to maintain operation at the desired high-performance, MHD-stable, plasma regime. Initial efforts towards current density profile control have led to the development of a control-oriented, physics-based, plasma-response model, which combines the magnetic diffusion equation with empirical correlations for the kinetic profiles and the non-inductive current sources. The developed control-oriented model has been successfully tailored to the NSTX-U geometry and actuators. Moreover, a series of efforts have been made towards the design of model-based controllers, including a linear-quadratic-integral optimal control strategy that can regulate the current density profile around a prescribed target profile while rejecting disturbances. In this work, the tracking performance of the proposed current-profile optimal controller is tested in numerical simulations based on the physics-oriented code TRANSP. These high-fidelity closed-loop simulations, which are a critical step before experimental implementation and testing, are enabled by a flexible framework recently developed to perform feedback control design and simulation in TRANSP.

Design and simulation of the snowflake divertor control for NSTX–U

This paper presents the development of a physics-based multiple-input-multiple-output algorithm for real-time feedback control of snowflake divertor (SFD) configurations on the National Spherical Torus eXperiment Upgrade (NSTX–U). A model of the SFD configuration response to applied voltages on the divertor control coils is first derived and then used, in conjunction with multivariable control synthesis techniques, to design an optimal state feedback controller for the configuration. To demonstrate the capabilities of the controller, a nonlinear simulator for axisymmetric shape control was developed for NSTX–U which simultaneously evolves the currents in poloidal field coils based upon a set of feedback-computed voltage commands, calculates the induced currents in passive conducting structures, and updates the plasma equilibrium by solving the free-boundary Grad–Shafranov problem. Closed-loop simulations demonstrate that the algorithm enables controlled operations in a variety of SFD configurations and provides capabilities for accurate tracking of time-dependent target trajectories for the divertor geometry. In particular, simulation results suggest that a time-varying controller which can properly account for the evolving SFD dynamical response is not only desirable but necessary for achieving acceptable control performance. The algorithm presented in this paper has been implemented in the NSTX–U Plasma Control System in preparation for future control and divertor physics experiments.

Full-f gyrokinetic simulation of turbulence in a helical open-field-line plasma

Curvature-driven turbulence in a helical open-field-line plasma is investigated using electrostatic five-dimensional gyrokinetic continuum simulations in an all-bad-curvature helical-slab geometry. Parameters for a National Spherical Torus Experiment scrape-off-layer plasma are used in the model. The formation and convective radial transport of plasma blobs is observed, and it is shown that the radial particle-transport levels are several times higher than diffusive Bohm-transport estimates. By reducing the strength of the poloidal magnetic field, the profile of the heat flux to the divertor plate is observed to broaden.

Effect of boronization on plasma-facing graphite surfaces and its correlation with the plasma behavior in NSTX-U

Boronization is a Plasma Facing Component (PFC) conditioning technique widely used in tokamak machines. The National Spherical Torus Experiment-Upgrade (NSTX-U) applied this conditioning, using a plasma glow with a deuterated Trimethyl-boron (d-TMB) and He mixture. The use of boronization during the campaign improved the plasma performance, allowing longer plasma discharges and H-mode access. The chemical state of an ATJ graphite sample, used as a proxy for the NSTX-U PFCs, was monitored in-situ using the Materials Analysis Particle Probe (MAPP) diagnostic and X-ray Photoelectron Spectroscopy (XPS). The XPS data showed a progressive rise (from < 5% to 23%) in the oxygen concentration of the boronized ATJ sample as the D+ fluence increased. Filterscopes were used to measure the light emitted by oxygen impurities in the plasma near the surface of the PFC. An increase in the registered magnitude of the OII line, normalized to the Dγ intensity, was observed as the concentration of O on the ATJ surface increased. The plasma performance was found to be strongly correlated to oxygen impurity concentrations at the plasma edge and on the PFC surface, as measured by the discharge length and access to the H-mode regime. In this work, we present a quantitative analysis of the evolution of the chemistry of the ATJ surface, and the oxygen presence in the plasma-material interface, and report relevant plasma parameters observed during the same period of time.

Supersonic Gas Injector for Plasma Fueling in the National Spherical Torus Experiment

A supersonic gas injector (SGI) has been developed for fueling and diagnostic applications on the National Spherical Torus Experiment (NSTX). It is comprised of a graphite converging-diverging Laval nozzle and a commercial piezoelectric gas valve mounted on a movable probe at a low-field-side midplane port location. Also mounted on the probe is a diagnostic package: a Langmuir probe, two thermocouples, and five pick-up coils for measuring toroidal, radial, vertical magnetic field components and magnetic fluctuations at the location of the SGI tip. The SGI flow rate is up to 33.25 Pa m3/ (1.75 × 1022 euterium particles/s), comparable to conventional NSTX gas injectors. The nozzle operates in a pulsed regime at room temperature and a reservoir gas pressure up to 665 kPa (5000 Torr). The deuterium jet Mach number of about 4 and the divergence half-angle of 5 to 25 deg have been measured in laboratory experiments simulating the NSTX environment. Reliable operation of the SGI and all mounted diagnostics at distances 0.01 to 0.20 m from the plasma separatrix has been demonstrated in NSTX experiments. The SGI has been used for fueling of ohmic and 2- to 4-MW neutral beam injection–heated L- and H-mode plasmas. Fueling efficiency in the range 0.1 to 0.3 has been obtained from the plasma electron inventory analysis. The SGI-fueling–based plasma discharge scenarios enabling better density control have been developed.

The high-k poloidal scattering system for NSTX-U.

An 8-channel, high-k poloidal far-infrared (FIR) scattering system is under development for the National Spherical Torus eXperiment Upgrade (NSTX-U). The 693 GHz poloidal scattering system replaces a 5-channel, 280 GHz high-k toroidal scattering system to study high-k electron density fluctuations on NSTX-U. The FIR probe beam launched from Bay G is aimed toward Bay L, where large aperture optics collect radiation at 8 simultaneous scattering angles ranging from 2° to 15°. The reduced wavelength in the poloidal system results in less refraction, and coupled with a new poloidal scattering geometry, extends measurement of poloidal wavenumbers from the previous limit of 7 cm-1 up to >40 cm-1. Steerable launch optics coupled with receiver optics that can be remotely translated in 5 axes allow the scattering volume to be placed from r/a = 0.1 out to the pedestal region (r/a ∼ 0.99) and allow for both upward and downward scattering to cover different regions of the 2D fluctuation spectrum.

Integrated plasma facing component calorimetry for measurement of shot integrated deposited energy in the NSTX-U.

The upgrade to the National Spherical Torus eXperiment (NSTX-U) [J. Menard et al., Nucl. Fusion 52, 083015 (2012)] increases the injected neutral beam power up to 12 MW and the plasma current up to Ip = 2 MA for plasma durations up to 5 s. The graphite plasma facing components have been re-designed to handle greater heat and energy fluxes than were seen in NSTX using a castellated design. We present the experimental testing and validation of a castellated graphite target, similar to the prototype tile design, instrumented with thermocouples at various depths in the castellation. During testing, incident heat flux is provided by a programmed electron beam system and surface temperatures are measured via infrared thermography directly viewing the target surface. It was found that the thermocouple response scaled linearly with the measured surface temperature rise regardless of thermocouple depth in the castellation. A sensitivity of 14.3 °C/kJ of deposited energy was found when treating individual castellations as a semi-infinite solid.

Orbit modeling of fast particle redistribution induced by sawtooth instability

Initial tests on the National Spherical Torus Experiment Upgrade (NSTX-U) (Menard et al 2012 Nucl. Fusion 083015) device suggest that introducing energy selectivity for sawtooth induced fast ion redistribution is required to improve the agreement between experimental and simulated quantities such as neutron rate and Fast-Ion D-Alpha profiles. The aim of this work is to assess the requirements to properly describe the behaviour of fast ions during a sawtooth crash for predictive sawtooth simulations. As the first step, in this work, we use the particle-following Orbit code to characterize the redistribution of fast particles. In order for a sawtooth crash to be simulated, a spatial and temporal displacement is implemented into the Orbit code. The perturbation amplitude is determined by comparison with experimental measurement of the neutron rate drop. The characteristics of fast ions with different orbit types are investigated in phase and real space. Due to a sawtooth crash, fast ion energy and angular momentum are modified resulting in the redistribution in phase space and orbit type change. The redistribution of fast ions in real space shows that the sawtooth instability brings different effect on fast particles with different orbit types as observed in experiments. The initial interpretative Transp simulation using the so-called kick model based on the Orbit modeling result shows an improvement of fast ion redistribution before and after a sawtooth crash but the neutron rate still has discrepancy compared to the experimental measurement.

NSTX-U advances in real-time deterministic PCIe-based internode communication

Distributing control mechanisms across modern commercial off the shelf (COTS) GNU/Linux systems often introduces difficult to mitigate non-deterministic behavior. Existing methods to address this problem involve non-real-time technologies such as Remote Direct Memory Access (RDMA) over Infiniband or custom Ethernet solutions, such as RDMA over Converged Ethernet (RoCE) or userspace Ethernet drivers, that trade determinism for ease of use or lower cost. The National Spherical Torus Experiment Upgrade (NSTX-U) is pursuing a new design that allows direct communication between heterogeneous systems with scalable, microsecond latency with 1 μs of jitter on that latency, outside of the constant transmission delay at the physical layer. The future design of the NSTX-U Real-time Communication System will utilize direct PCIe-to-PCIe communication with kernel support tuned for low overhead, allowing two (or more, through a switch) real-time (RT) systems to communicate and share resources as one larger entity. This greatly increases the processing capability of the primary Plasma Control System (PCS), turning previously insurmountable computational challenges into a more manageable divide and conquer parallel task.

On the scattering correction of fast-ion D-alpha signals on NSTX-U.

Analysis of fast-ion D-alpha (FIDA) data on National Spherical Torus Experiment-Upgrade (NSTX-U) shows that the cold Dα line contaminates the FIDA baseline. The scattered light is comparable to the FIDA emission. A scattering correction is required to extract the FIDA signal. Two methods that relate the scattered light contamination to the intensity of the cold Dα line are employed. One method uses laboratory measurements with a calibration lamp; the other method uses data acquired during plasma operation and singular value decomposition analysis. After correction, both the FIDA spectra and spatial profile are in better agreement with theoretical predictions.

Regarding the optimization of O1-mode ECRH and the feasibility of EBW startup on NSTX-U

Recently published scenarios for fully non-inductive startup and operation on the National Spherical Torus eXperiment Upgrade (NSTX-U) (Menard et al 2012 Nucl. Fusion 52 083015) show Electron Cyclotron Resonance Heating (ECRH) as an important component in preparing a target plasma for efficient High Harmonic Fast Wave and Neutral Beam heating. The modeling of the propagation and absorption of EC waves in the evolving plasma is required to define the most effective window of operation, and to optimize the launcher geometry for maximal heating and current drive during this window. Here, we extend a previous optimization of O1-mode ECRH on NSTX-U to account for the full time-dependent performance of the ECRH using simulations performed with TRANSP. We find that the evolution of the density profile has a prominent role in the optimization by defining the time window of operation, which in certain cases may be a more important metric to compare launcher performance than the average power absorption. This feature cannot be captured by analysis on static profiles, and should be accounted for when optimizing ECRH on any device that operates near the cutoff density. Additionally, the utility of the electron Bernstein wave (EBW) in driving current and generating closed flux surfaces in the early startup phase has been demonstrated on a number of devices. Using standalone GENRAY simulations, we find that efficient EBW current drive is possible on NSTX-U if the injection angle is shifted below the midplane and aimed towards the top half of the vacuum vessel. However, collisional damping of the EBW is projected to be significant, in some cases accounting for up to 97% of the absorbed EBW power.

Scenario development during commissioning operations on the National Spherical Torus Experiment Upgrade

The National Spherical Torus Experiment Upgrade (NSTX-U) will advance the physics basis required for achieving steady-state, high-beta, and high-confinement conditions in a tokamak by accessing high toroidal fields (1 T) and plasma currents (1.0–2.0 MA) in a low aspect ratio geometry (A = 1.6–1.8) with flexible auxiliary heating systems (12 MW NBI, 6 MW HHFW). This paper describes the progress in the development of L- and H-mode discharge scenarios and the commissioning of operational tools in the first ten weeks of operation that enable the scientific mission of NSTX-U. Vacuum field calculations completed prior to operations supported the rapid development and optimization of inductive breakdown at different values of ohmic solenoid current. The toroidal magnetic field (BT0 = 0.65 T) exceeded the maximum values achieved on NSTX and novel long-pulse L-mode discharges with regular sawtooth activity exceeded the longest pulses produced on NSTX (tpulse > 1.8 s). The increased flux of the central solenoid facilitated the development of stationary L-mode discharges over a range of density and plasma current (Ip). H-mode discharges achieved similar levels of stored energy, confinement (H98y,2 > 1) and stability (βN/βN-nowall > 1) compared to NSTX discharges for Ip ⩽ 1 MA. High-performance H-mode scenarios require an L–H transition early in the Ip ramp-up phase in order to obtain low internal inductance (li) throughout the discharge, which is conducive to maintaining vertical stability at high elongation (κ > 2.2) and achieving long periods of MHD quiescent operations. The rapid progress in developing L- and H-mode scenarios in support of the scientific program was enabled by advances in real-time plasma control, efficient error field identification and correction, effective conditioning of the graphite wall and excellent diagnostic availability.

Electron-density-sensitive Line Ratios of Fe xiii– xvi from Laboratory Sources Compared to CHIANTI

Abstract 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.

Plasma boundary shape control and real-time equilibrium reconstruction on NSTX-U

The upgrade to the National Spherical Torus eXperiment (NSTX-U) included two main improvements: a larger center-stack, enabling higher toroidal field and longer pulse duration, and the addition of three new tangentially aimed neutral beam sources, which increase available heating and current drive, and allow for flexibility in shaping power, torque, current, and particle deposition profiles. To best use these new capabilities and meet the high-performance operational goals of NSTX-U, major upgrades to the NSTX-U control system (NCS) hardware and software have been made. Several control algorithms, including those used for real-time equilibrium reconstruction and shape control, have been upgraded to improve and extend plasma control capabilities. As part of the commissioning phase of first plasma operations, the shape control system was tuned to control the boundary in both inner-wall limited and diverted discharges. It has been used to accurately track the requested evolution of the boundary (including the size of the inner gap between the plasma and central solenoid, which is a challenge for the ST configuration), X-point locations, and strike point locations, enabling repeatable discharge evolutions for scenario development and diagnostic commissioning.

Measurement of the passive fast-ion D-alpha emission on the NSTX-U tokamak

On National Spherical Torus Experiment Upgrade, the passive fast-ion D-alpha (passive-FIDA) spectra from charge exchange (CX) between the beam ions and the background neutrals are measured and simulated. The results indicate that the passive-FIDA signal is measurable and comparable to the active-FIDA on several channels, such as at the major radius R = 117 cm. Here, active-FIDA means the active D-alpha emission from the fast ions that CX with the injected neutrals. The shapes of measured spectra are in agreement with FIDASIM simulations on many fibers. Furthermore, the passive-FIDA spatial profile agrees with the simulation. When making measurements of active-FIDA in the edge region using time-slice subtraction, variations in the passive-FIDA contribution to the signal should be considered.

Density perturbation mode structure of high frequency compressional and global Alfvén eigenmodes in the National Spherical Torus Experiment using a novel reflectometer analysis technique

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).

Drift kinetic effects on plasma response in high beta spherical tokamak experiments

The high β plasma response to rotating external magnetic perturbations is numerically studied and compared with the National Spherical Torus Experiment (NSTX). The hybrid magnetohydrodynamic(MHD)-kinetic modeling shows that drift kinetic effects are important in resolving the disagreement of plasma response between the ideal MHD prediction and the NSTX experimental observation when plasma pressure reaches and exceeds the no-wall limit (Troyon et al 1984 Plasma Phys. Control. Fusion 26 209). Since the external rotating fields and high plasma rotation are presented in the NSTX experiments, the importance of the resistive wall effect and plasma rotation in determining the plasma response is also identified, where the resistive wall suppresses the plasma response through the wall eddy current. The inertial energy due to plasma rotation destabilizes the plasma. The complexity of the plasma response in this study indicates that MHD modeling, including comprehensive physics, e.g. the drift kinetic effects, resistive wall and plasma rotation, are essential in order to reliably predict the plasma behavior in a high beta spherical tokamak device.