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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Impact of ELM filaments on divertor heat flux dynamics in NSTX

In this study, the ELM induced change in wetted area (A wet ) and peak heat flux (q peak ) of divertor heat flux is investigated as a function of the number of striations, which represent ELM filaments, observed in the heat flux profile in NSTX. More striations are found to lead to larger A wet and lower q peak . The typical number of striations observed in NSTX is 0–9, while 10–15 striations are normally observed in other machines such as JET, and the ELM contracts heat flux profile when the number of striations is less than 3–4 but broadens it with more of them. The smaller number of striations in NSTX is attributed to the fact that NSTX ELMs are against kink/peeling boundary with lower toroidal mode number (n = 1–5), while typical peeling–ballooning ELMs have higher mode number of n = 10–20. For ELMs with smaller number of striations, relative A wet change is rather constant and qpeak change rapidly increases with increasing ELM size, while A wet change slightly increases leading to a weaker increase of q peak change for ELMs with larger number of striations, both of which are unfavourable trend for the material integrity of divertor tiles.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Microtearing instabilities and electron thermal transport in low and high collisionality NSTX discharges

Microtearing mode (MTM) real frequency, growth rate, magnetic fluctuation amplitude and resulting electron thermal transport are studied in systematic NSTX scans of relevant plasma parameters. Here, the dependency of the MTM real frequency and growth rate on plasma parameters, suitable for low and high collision NSTX discharges, is obtained by using the reduced MTM transport model. The plasma parameter dependencies are compared and found to be consistent with the results obtained from MTM using the Gyrokinetic GYRO code. The scaling trend of collision frequency and plasma beta is found to be consistent with the global energy confinement trend observed in the NSTX experiment. The strength of the magnetic fluctuation is found to be consistent with the gyrokinetic estimate. In earlier studies, it was found that the version of the Multi-Mode (MM) anomalous transport model, which did not contain the effect of MTMs, provided an appropriate description of the electron temperature profiles in standard tokamak discharges and not in spherical tokamaks. When the MM model, which involves transport associated with MTMs, is incorporated in the TRANSP code and is used in the study of electron thermal transport in NSTX discharges, it is observed that the agreement with the experimental electron temperature profile is substantially improved.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

On the frequency bifurcations of the MHD startup modes in NSTX

The observed bifurcations of the low frequency (<50 kHz) and low toroidal periodicity (n < 5) magnetohydrodynamic (MHD) activity often present in the initial part of the National Spherical Tokamak Experiment (NSTX) discharges can be explained by the evolution of the radial profile of the safety factor (q=rB φ /RB θ ) crossing multiple rational surfaces in the core. Important performance limiting instability mechanisms in the NSTX spherical tokamak are often linked to low frequency and low-n MHD activity. They are quite common in long-pulse NSTX plasmas. They can be present at the beginning of the plasma current flat-top, at the end of the discharge or during the whole duration, and they have been observed to deleteriously impact performance over a wide range of q 95 . An interesting feature observed in some NSTX discharges is the presence of a bifurcation in the frequency of the low n modes, as low as n = 1, that have frequencies comparable to the plasma core rotation divided by n. Equilibrium reconstructions constrained by magnetic diagnostics data and motional stark effect pitch angle radial profiles suggest that the observed bifurcations are linked to a fast evolving minimum value of q. Finally, 3D non-linear resistive MHD simulations show that these modes are ideal and exist as non-resonant before the correspondent rational surface enters the plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Analysis of the impact of parallel magnetic fluctuations on linear gyrokinetic stability in NSTX-U and verification of gyro-fluid models

In this work, we use the CGYRO gyrokinetic code to analyze two L- and one H-mode discharges from the National Spherical Torus Experiment (NSTX) and NSTX-Upgrade (NSTX-U) selected due to their different mix of ion-scale driftwaves, ion temperature gradient (ITG) mode and trapped electron mode (TEM), and electromagnetic instabilities, kinetic ballooning mode (KBM), and micro-tearing mode (MTM) in the plasma core. It is found that the effect of parallel magnetic fluctuations is strongly destabilizing to the unstable KBMs compared to calculations with only perpendicular magnetic fluctuations. Two discharges have a mix of ITG/TEM and MTMs that are predicted to be dominant instability across the plasma radius. The parallel magnetic fluctuations are found to have little effect on the MTM stability but are destabilizing to ITG/TEM modes. To test the validity of the gyro-fluid linear stability codes TGLF and GFS at low aspect ratio, a database of linear growth rates has been created using the CGYRO gyrokinetic code. The database is comprised of various parameter scans around a standardized set of NSTX-U core parameters. It contains a group of electrostatic cases and an electromagnetic group that includes the effects of perpendicular and parallel magnetic fluctuations. Comparing the results from the GFS and TGLF models, we find that GFS exhibits the best agreement with the database of CGYRO linear growth rates. Comparing the model results for the electromagnetic scans shows that GFS captures the effects of parallel magnetic fluctuations accurately, while the TGLF model does not, as it lacks sufficient perpendicular energy resolution.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Edge loss of high-harmonic fast-wave heating power in NSTX: a cylindrical model

Efficient high-harmonic fast-wave (HHFW) heating in the National Spherical Torus Experiment Upgrade (NSTX-U) would facilitate experiments in turbulence, transport, fast-ion studies, and more. However, previous HHFW operation in NSTX exhibited a large loss of fast-wave power to the divertor along the scrape-off layer field lines for edge densities above the fast-wave cutoff. It was postulated that the wave amplitude is enhanced in the scrapeoff layer due to cavity-like modes, and that these enhanced fields drive sheath losses through RF rectification. As part of ongoing work to confirm this hypothesis, we have developed a cylindrical cold-plasma model to identify and understand scenarios where a substantial fraction of wave power is confined to the plasma periphery. We previously identified a peculiar class of modes, named annulus resonances, that conduct approximately half of their wave power in the periphery and can also account for a significant fraction of the total wave power. Here, we study the influence of annulus resonances on wave field reconstructions and find instances where annulus-resonant modes dominate the spectrum and trap over half of the total wave power at the edge. The work is part of an ongoing effort to determine the mechanism underlying these scrape-off layer losses in NSTX, identify optimal conditions for operation in NSTX-U, and predict whether similar losses occur for the ion-cyclotron minority heating scheme for both current experiments and future devices such as ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Reversed magnetic shear scenario development in NSTX-U using TRANSP

Abstract 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-( q min = 1) MHD events, requires maintaining a large plasma volume, and increasing the core T e , either via increased plasma current and/or adding heating power. Using this procedure, RMS was sustained for ∼1 s, with q min > 1 for that period.

Galante, M. E. (ORCID:0009000098149425)↗
DOE OSTI2024

Real-Time Ethernet Interface for NSTX-U’s Thomson Scattering Diagnostic (2023)

Here, 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.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Microtearing Instabilities and Electron Thermal Transport in Low and High Collisionality NSTX Discharges

Microtearing mode (MTM) real frequency, growth rate, magnetic fluctuation amplitude and resulting electron thermal transport are studied in systematic NSTX scans of relevant plasma parameters. The dependency of the MTM real frequency and growth rate on plasma parameters, suitable for low and high collision NSTX discharges, is obtained by using the reduced MTM transport model [T. Rafiq, et al., Phys. Plasmas 23, 062507 (2016)]. The plasma parameter dependencies are compared and found to be consistent with the results obtained from MTM using the Gyrokinetic GYRO code. The scaling trend of collision frequency and plasma beta is found to be consistent with the global energy confinement trend observed in the NSTX experiment. The strength of the magnetic fluctuation is found to be consistent with the gyrokinetic estimate.In earlier studies, it was found that the version of the Multi-Mode (MM) anomalous transport model, which did not contain the effect of MTMs, provided an appropriate description of the electron temperature profiles in standard tokamak discharges and not in spherical tokamaks. When the MM model, which involves transport associated with MTMs, is incorporated in the TRANSP code and is used in the study of electron thermal transport in NSTX discharges, it is observed that the agreement with the experimental electron temperature profile is substantially improved.

fusion plasma technology↗
DOE OSTI2012

Edge Plasma Simulations in NSTX and CTF: Synergy of Lithium Coating, Non-Diffusive Anomalous Transport and Drifts

This is the final report for the Research Grant DE-FG02-08ER54989 'Edge Plasma Simulations in NSTX and CTF: Synergy of Lithium Coating, Non-Diffusive Anomalous Transport and Drifts'. The UCSD group including: A.Yu. Pigarov (PI), S.I. Krasheninnikov and R.D. Smirnov, was working on modeling of the impact of lithium coatings on edge plasma parameters in NSTX with the multi-species multi-fluid code UEDGE. The work was conducted in the following main areas: (i) improvements of UEDGE model for plasma-lithium interactions, (ii) understanding the physics of low-recycling divertor regime in NSTX caused by lithium pumping, (iii) study of synergistic effects with lithium coatings and non-diffusive ballooning-like cross-field transport, (iv) simulation of experimental multi-diagnostic data on edge plasma with lithium pumping in NSTX via self-consistent modeling of D-Li-C plasma with UEDGE, and (v) working-gas balance analysis. The accomplishments in these areas are given in the corresponding subsections in Section 2. Publications and presentations made under the Grant are listed in Section 3.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Propagation And Damping Of High Harmonic Fast Waves And Electron Cyclotron Waves In The Nstx-U-Device

For the spherical torus and for conventional tokamaks to be viable means of producing fusion energy there are two important requirements. The first is to heat the plasma in the experimental devices to high temperatures; preferably, to fusion relevant temperatures. The second is to drive a current in the plasma so that the fusion device operates in a steady-state mode. A pulsed fusion reactor for energy production is not desirable. A favored means to achieve both of these requirements is through the use of microwaves, commonly referred to as radio frequency waves. This proposal was to understand observations from the National Spherical Torus Experiment Upgrade (NSTX-U) device on radio frequency wave heating of the confined plasma during operations with high harmonic fast waves. The proposed research was also directed toward assisting in the optimization of the experimental modes of operation so as to maximize the heating efficiency. A detailed analysis of the experiments has shown that about one-third of the input radio frequency power does not make it into the core region of the plasma where heating is required. The losses are believed to occur primarily in the edge region where the radio frequency power is initially coupled into the plasma. Using sophisticated computational codes and theoretical models, that have been developed for the NSTX-U experimental program, this research quantified the physical processes which occur in the edge plasma and lead to a decrease in the heating efficiency. The effect of edge turbulence on the scattering of radio frequency waves was studied for waves in the ion cyclotron and electron cyclotron range of frequencies. For high harmonic fast waves in the ion cyclotron range of frequencies, simulations were carried out using a full-wave solver that was extended to the vacuum vessel wall. These simulations successfully reproduced trends that were observed experimentally. This simulation capability was also used to study the interaction of high harmonic fast waves with the scrape-off layer plasma. More recently, a new technique has been developed for the calculation of radio frequency waves in toroidal geometry that enables the simultaneous incorporation of antenna geometry, plasma facing components, the scrape off-layer, and core propagation in the high harmonic fast wave regime. The effect of turbulent plasmas on the mode conversion of extraordinary electromagnetic waves in the electron cyclotron range of frequencies in the NSTX-U plasma has been studied theoretically. In the over-dense plasmas of NSTX-U, the traditional electromagnetic waves are not effective for delivering energy and momentum to the core plasma. However, electron Bernstein waves generated by mode conversion are efficient carriers of wave energy and momentum to the core. The mode conversion occurs in the turbulent edge region. A theory for mode conversion in such plasmas has been developed. The parameters that can optimize the mode conversion can be readily determined. Also, the model is useful for experimentalists to analyze observation on a shot-to-shot basis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Propagation And Damping Of High Harmonic Fast Waves And Electron Cyclotron Waves In The Nstx-U-Device

For the spherical torus and for conventional tokamaks to be viable means of producing fusion energy there are two important requirements. The first is to heat the plasma in the experimental devices to high temperatures; preferably, to fusion relevant temperatures. The second is to drive a current in the plasma so that the fusion device operates in a steady-state mode. A pulsed fusion reactor for energy production is not desirable. A favored means to achieve both of these requirements is through the use of microwaves, commonly referred to as radio frequency waves. This proposal was to understand observations from the National Spherical Torus Experiment Upgrade (NSTX-U) device on radio frequency wave heating of the confined plasma during operations with high harmonic fast waves. The proposed research was also directed toward assisting in the optimization of the experimental modes of operation so as to maximize the heating efficiency. A detailed analysis of the experiments has shown that about one-third of the input radio frequency power does not make it into the core region of the plasma where heating is required. The losses are believed to occur primarily in the edge region where the radio frequency power is initially coupled into the plasma. Using sophisticated computational codes and theoretical models, that have been developed for the NSTX-U experimental program, this research quantified the physical processes which occur in the edge plasma and lead to a decrease in the heating efficiency. The effect of edge turbulence on the scattering of radio frequency waves was studied for waves in the ion cyclotron and electron cyclotron range of frequencies. For high harmonic fast waves in the ion cyclotron range of frequencies, simulations were carried out using a full-wave solver that was extended to the vacuum vessel wall. These simulations successfully reproduced trends that were observed experimentally. This simulation capability was also used to study the interaction of high harmonic fast waves with the scrape-off layer plasma. More recently, a new technique has been developed for the calculation of radio frequency waves in toroidal geometry that enables the simultaneous incorporation of antenna geometry, plasma facing components, the scrape off-layer, and core propagation in the high harmonic fast wave regime. The effect of turbulent plasmas on the mode conversion of extraordinary electromagnetic waves in the electron cyclotron range of frequencies in the NSTX-U plasma has been studied theoretically. In the over-dense plasmas of NSTX-U, the traditional electromagnetic waves are not effective for delivering energy and momentum to the core plasma. However, electron Bernstein waves generated by mode conversion are efficient carriers of wave energy and momentum to the core. The mode conversion occurs in the turbulent edge region. A theory for mode conversion in such plasmas has been developed. The parameters that can optimize the mode conversion can be readily determined. Also, the model is useful for experimentalists to analyze observation on a shot-to-shot basis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Test of Point-Source Helicity Injection for Non-Solenoidal Startup in NSTX

This grant supported research activity focused on the development of Local Helicity Injection (LHI) as a technique for initiation and growth of spherical tokamak plasmas without the use of a traditional ohmic solenoid for magnetic induction. The LHI technique consists of injection of high-density current streams into a vacuum magnetic field consisting of toroidal and vertical fields. Under appropriate conditions, these streams become unstable and support magnetic reconnection to allow the system to reconfigure itself into a tokamak-like plasma magnetic configuration. This research program focused on three aspects of this plasma startup technique: 1) development and tests of appropriate electron current sources that provide sufficiently high injection currents into the near-plasma edge without strongly contaminating the resultant tokamak plasma with impurities; 2) development and validation of a 0-D power balance model to describe the evolution of the plasma current I p (t) during this startup process; and 3) initial estimates of I p levels achievable with this technique in NSTX-U scale plasmas. The 0-D model development included extremely low aspect ratio corrections to analytical formulae for the plasma inductances that produce voltages due to rapid changes in plasma size and location. Applications of this model to sample plasmas in the Pegasus ultra-low aspect ratio tokamak confirms the expectation that I p (t) is dominated by these induced voltages when injection occurs from the outer low-field side of the plasma. Conversly, I p (t) is dominated by an effective toroidal loop voltage arising from helicity conservation when injectors are placed in the inner high-field side divertor region. The activities pursued here, combined with more general research on LHI followed by the Pegasus experiment group, provide the tools needed for initial consideration of a MA-level plasma startup capability for the NSTX-U experiment. Simplified modeling for LHI on NSTX-U indicates that no showstoppers exist, and ~ 0.8MA appears accessible using a moderate current injection facility. However, many issues remain to be resolved to address extrapolating this technique to the larger sizes and magnetic field strengths of NSTX-U and larger fusion-grade tokamaks. These include, among others: I p scaling with helicity drive input; extending current injector technology; understanding electron confinement and the resulting helicity dissipation processes that determine the achievable I p ; the role of plasma stability in controlling the current evolution; and controlling plasma-material interactions and impurity content. These open issues will be further explored in a future experiment dedicated to integrated studies of non-solenoidal startup at higher magnetic fields and plasma currents.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Plasma shape and position control development for NSTX-U using the GSEvolve plasma simulator

NSTX-U discharges have been reproduced using the GSEvolve plasma simulator in order to improve their control and reduce oscillations during the plasma current (Ip) ramp-up phase, and with the ultimate goal of enabling reliable H-mode access with a double-null shape at high elongation, low internal inductance, and high plasma performance. The control objective is to eliminate undesired oscillations in the plasma shape and vertical position that start when the plasma diverts during the Ip ramp-up phase, and that happened recurrently during the first NSTX-U experimental campaign. Model-based control solutions have been developed, including alternative H-mode access recipes and a Multi-Input Multi-Output (MIMO) decoupling controller. These solutions have demonstrated their capability to reduce or eliminate the aforementioned oscillations in GSEvolve simulations, enabling reproducible H-mode access. The results of this work aim to inform NSTX-U experimental operations on potential control developments to maximize its scientific output, and highlight the powerful capabilities of high-fidelity control-simulation frameworks like GSEvolve.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Prospects for the Thomson scattering system on NSTX-Upgrade

The paper discusses the projected configuration of the Thomson system on the National Spherical Torus Experiment (NSTX-U). In this paper, we discuss the projected configuration of the Thomson system on NSTX-U. More specifically, we determine, through both optical modeling of the collection optics and in-vessel measurements, that the collecting fibers are to be displaced by at most 1 cm toward the imaging plane along the optical axis. Finally, we estimate the performance of the Thomson system in measuring the electron temperature for NSTX-U discharges.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Towards identifying the mechanisms underlying field-aligned edge-loss of HHFW power on NSTX

Fast-wave heating will be a major heating scheme on ITER, as it can heat ions directly and is relatively unaffected by the large machine size unlike neutral beams. However, fast-wave interactions with the plasma edge can lead to deleterious effects such as, in the case of the high-harmonic fast-wave (HHFW) system on NSTX, large losses of fast-wave power in the scrape off layer (SOL) under certain conditions. In such scenarios, a large fraction of the lost HHFW power is deposited on the upper and lower divertors in bright spiral shapes. The responsible mechanism(s) has not yet been identified but may include fast-wave propagation in the scrape off layer, parametric decay instability, and RF currents driven by the antenna reactive fields. Understanding and mitigating these losses is important not only for improving the heating and current-drive on NSTX-Upgrade but also for understanding fast-wave propagation across the SOL in any fast-wave system. This talk summarizes experimental results demonstrating that the flow of lost HHFW power to the divertor regions largely follows the open SOL magnetic field lines. This lost power flux is relatively large close to both the antenna and the last closed flux surface with a reduced level in between, so the loss mechanism cannot be localized to the antenna. At the same time, significant losses also occur along field lines connected to the inboard edge of the bottom antenna plate. The power lost within the spirals is roughly estimated, showing that these field-aligned losses to the divertor are significant but may not account for the total HHFW loss. To elucidate the role of the onset layer for perpendicular fast-wave propagation with regards to fast-wave propagation in the SOL, a cylindrical cold-plasma model is being developed. This model, in addition to advanced RF codes such as TORIC and AORSA, is aimed at identifying the underlying mechanism(s) behind these SOL losses, to minimize their effects in NSTX-U, and to predict their importance in ITER.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

CQL3D-Hybrid-FOW modeling of the temporal dynamics of NSTX NBI+HHFW discharges

The CQL3D Fokker-Planck code[1] has been upgraded to include physics of finite-orbit-width (FOW) guiding-center orbits[2,3], as compared with the previous zero-orbit-width (ZOW) model, and a recent first-order orbit calculation[2]. The Fast Ion Diagnostic FIDA[4,5] signal resulting from neutral beam (NBI) and high harmonic fast wave (HHFW) RF power injected into the NSTX spherical tokamak can now be modeled quite accurately, using ion distributions from the CQL3D-Hybrid-FOW code, a rapidly executing variant that includes FOW+gyro-orbit losses to the plasma edge, FOW effects on NBI injection and HHFW diffusion, but does not include neoclassical radial diffusion. Accurate simulation of prompt fast ion (FI) losses is a key feature of the marked modeling improvement relative to previous ZOW results. By comparing NBI-only and NBI+HHFW shots, independent confirmation of the usual 35% edge loss of HHFW in NSTX is obtained. Further, HHFW prompt losses from the plasma core are shown to be 3X as large (>25%) as the NBI-only case. The modulated NBI and time-dependent background plasma variations and charge exchange losses of fast ions are accounted for, and the temporal neutron variation is in approximate agreement with NSTX observations.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2019

Overview of the Systems Engineering Process as It Applies to The National Spherical Torus Experiment - Upgrade (NSTX-U)

This paper addresses the systems engineering (SE) processes used for the National Spherical Torus Experiment-Upgrade (NSTX-U). It focuses on SE across the life cycle of the system, including requirements management, interface control, risk management, integration, and verification/validation. This is particularly significant as NSTX-U includes new systems and an existing plant and reused systems from past projects such as the Tokamak Fusion Test Reactor (TFTR). The implementation of SE provides the ability to control complexity, improve communications, identify risks early, and prevent defects. Systems engineering principles are applied to enhance the integration while maintaining relevance in plasma research. Furthermore, these principles define a deliberate process to identify and resolve issues early in the development cycle, thus reducing risks and optimizing outputs. They also establish relationships to gather knowledge from experts and stakeholders, supporting the continued ability of NSTX-U in building and maintaining an operational system able to adapt to changing environments and emerging requirements.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Initial operation of the NSTX-Upgrade real-time velocity diagnostic

A real-time velocity (RTV) diagnostic based on active charge-exchange recombination spectroscopy is now operational on the National Spherical Torus Experiment-Upgrade (NSTX-U) spherical torus (Menard et al 2012 Nucl. Fusion 52 083015). We designed the system in order to supply plasma velocity data in real time to the NSTX-U plasma control system, as required for the implementation of toroidal rotation control. Our measurements are available from four radii at a maximum sampling frequency of 5 kHz. Post-discharge analysis of RTV data provides additional information on ion temperature, toroidal velocity and density of carbon impurities. Furthermore, examples of physics studies enabled by RTV measurements from initial operations of NSTX-U are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗