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DOE OSTI2016

Modeling and control of plasma rotation for NSTX using neoclassical toroidal viscosity and neutral beam injection

A model-based feedback system is presented to control plasma rotation in a magnetically confined toroidal fusion device, to maintain plasma stability for long-pulse operation. This research uses experimental measurements from the National Spherical Torus Experiment (NSTX) and is aimed at controlling plasma rotation using two different types of actuation: momentum from injected neutral beams and neoclassical toroidal viscosity generated by three-dimensional applied magnetic fields. Based on the data-driven model obtained, a feedback controller is designed, and predictive simulations using the TRANSP plasma transport code show that the controller is able to attain desired plasma rotation profiles given practical constraints on the actuators and the available measurements of rotation.

feedback control↗

DOE OSTI2024

NSTX-U research advancing the physics of spherical tokamaks

Abstract The objectives of NSTX-U research are to reinforce the advantages of STs while addressing the challenges. To extend confinement physics of low- A , high beta plasmas to lower collisionality levels, understanding of the transport mechanisms that set confinement performance and pedestal profiles is being advanced through gyrokinetic simulations, reduced model development, and comparison to NSTX experiment, as well as improved simulation of RF heating. To develop stable non-inductive scenarios needed for steady-state operation, various performance-limiting modes of instability were studied, including MHD, tearing modes, and energetic particle instabilities. Predictive tools were developed, covering disruptions, runaway electrons, equilibrium reconstruction, and control tools. To develop power and particle handling techniques to optimize plasma exhaust in high performance scenarios, innovative lithium-based solutions are being developed to handle the very high heat flux levels that the increased heating power and compact geometry of NSTX-U will produce, and will be seen in future STs. Predictive capabilities accounting for plasma phenomena, like edge harmonic oscillations, ELMs, and blobs, are being tested and improved. In these ways, NSTX-U researchers are advancing the physics understanding of ST plasmas to maximize the benefit that will be gained from further NSTX-U experiments and to increase confidence in projections to future devices.

74 ATOMIC AND MOLECULAR PHYSICS↗

DOE OSTI2024

NSTX-U National Research Program: White Paper in Response to Call from FESAC Sub-Committee

Both scientific and technical innovation is needed for the realization of an attractive engineering solution for a timely and cost-effective Pilot Plant, the design and construction of which is the overarching recommendation of the FESAC Long Range Plan, and the 2021 NASEM Pilot Plant reports, which underpin the Bold Decadal Vision. The two most significant plasma physics gaps to close for a Compact Pilot Plant (CPP) are core confinement improvement and heat flux mitigation, neither of which have been closed in an integrated fashion for any planned fusion power production device. High core confinement and stability are essential for producing majority self-driven plasmas in CPPs with reduced size and auxiliary heating power requirements, with an improvement in confinement being the major driver for cost reduction of a CPP. The National Spherical Tokamak Experiment - Upgrade (NSTX-U) is a unique low aspect ratio research facility that will address the fundamental challenge of developing the science and technology basis for a CPP design that integrates high core and edge confinement with the ability to mitigate very high incident heat fluxes. NSTX-U capabilities will enable the high performance, already achieved on NSTX, to extend into physics regimes much closer to those anticipated in Spherical Tokamak (ST)-based CPPs. These confinement and stability properties will be assessed by a full complement of diagnostics and analysis tools, which will also aid in the development of the underlying theory and predictive models needed for further optimization. Both conventional and transformative heat flux mitigation methods, such as liquid lithium plasma-facing components, will be developed and tested in-situ in NSTX-U at incident heat fluxes of ~100 MW/m 2 , and will inform plans and reduce risk for a subsequent major upgrade to the device to fully heated, high-Z wall and full liquid lithium divertor capability, a technology that potentially could then be implemented on any magnetic confinement device at any aspect ratio. NSTX-U research is fully complementary to programs performed on other STs, nationally and internationally. Furthermore, NSTX-U research has a direct connection to the private sector by informing design choices for future power production facilities being developed by these companies. The NSTX-U program will operate as a national User Facility, with collaborating researchers, engineers, and graduate students from 19 outside institutions, and open to participation and experiments led by researchers from both public and private entities. The research program will advance workforce development through training of young scientists, engineers, and technicians, and it will also serve for further diagnostic innovation, especially for high heat flux and high-Z wall environments, and implementation of advanced artificial intelligence (AI) for plasma and heat flux control.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Quantum-Classical Science for the Plasma-Material Interface in NSTXU (Final Technical Report)

Plasma facing surfaces (PFS) have a profound influence on plasma performance and choice of material for PFSs is a burning issue for tokamak reactors. Using lithium wall conditioning in NSTX has resulted in a beneficial influence on plasma behavior including control of hydrogen recycling and improvements in energy confinement, longer confinement times, edge density control, and the elimination of Edge Localized Modes. A major challenge for the NSTX-U (increased power flux, increased campaign time) as well as for other magnetic fusion devices is to extend high-performance plasmas for very long durations, and to integrate this high performance with plasma facing components (PFCs) that can withstand very high heat and particle fluxes while maintaining structural integrity with minimal retention of fusion fuel. This project directly contributed to filling up the knowledge gaps and data for the dynamics of the plasma-facing interfaces in the NSTX-U at nano and nano-to-micro scales. This was achieved by both deciphering the poorly understood nanoscale phenomenology of plasma-surface interactions and providing the connection and comparison with the experiments which provide understanding at the mesoscopic scales. We modelled the dynamic response and evolution of a mixed material surfaces (D, Li, C, B, O) to bombardment by plasma particles, and investigated microstructure changes, erosion, surface chemistry, deuterium implantation and permeation. The main objectives of the project are i) a comparison of Li and B effects when deposited on carbon, and ii) the role of oxygen and other impurities (e.g. carbon, boron) in the lithium performance. In addition to predicting and understanding the phenomenology of the processes, the project achieved deliverables on the i) plasma induced erosion rates of PFCs (outer wall, divertor, etc.), including chemical and physical sputtering yields at various temperatures (300-700K), ii) deuterium uptake/recycling rates, and iii) accurate probability rates for the processes and reactions in i-ii which were used to understand the experimental results of the J.P. Allain and B. Koel NSTX-U projects. Lithium was a distinctive focus of the project. It easily charges in contact with other fusion material atoms, giving partially or fully its external electron cloud to other atoms in its vicinity and polarizing the mixed material. Our research strategy used experimentally validated multiscale theory applied to the large, multicomponent material surface systems. With the use of US supercomputing facilities we simulated evolution of atomic charges, chemistry, retention of deuterium, sputtering and morphology by introducing the corrections to the classical molecular dynamics, like are electronegativity equalization method (EEM) and quantum-classical molecular dynamics (QCMD). The goal was to understand and quantify the dynamics of the creation and evolution of the plasma-material interface under irradiation by atoms and molecules at boronized carbon, lithiated carbon, and boronized-lithiated carbon for both solid and self-healing liquid-metal divertor.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Real-time radiative divertor feedback control development for the NSTX-U tokamak using a vacuum ultraviolet spectrometer

A radiative divertor technique is planned for the NSTX-U tokamak to prevent excessive erosion and thermal damage of divertor plasma-facing components in H-mode plasma discharges with auxiliary heating up to 12 MW. In the radiative (partially detached) divertor, extrinsically seeded deuterium or impurity gases are used to increase plasma volumetric power and momentum losses. A real-time feedback control of the gas seeding rate is planned for discharges of up to 5 s duration. The outer divertor leg plasma electron temperature T{sub e} estimated spectroscopically in real time will be used as a control parameter. A vacuum ultraviolet spectrometer McPherson Model 251 with a fast charged-coupled device detector is developed for temperature monitoring between 5 and 30 eV, based on the Δn = 0, 1 line intensity ratios of carbon, nitrogen, or neon ion lines in the spectral range 300–1600 Å. A collisional-radiative model-based line intensity ratio will be used for relative calibration. A real-time T{sub e}-dependent signal within a characteristic divertor detachment equilibration time of ∼10–15 ms is expected.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2017

Recent progress in understanding electron thermal transport in NSTX

The anomalous level of electron thermal transport inferred in magnetically confined configurations is one of the most challenging problems for the ultimate realization of fusion power using toroidal devices: tokamaks, spherical tori and stellarators. It is generally believed that plasma instabilities driven by the abundant free energy in fusion plasmas are responsible for the electron thermal transport. The National Spherical Torus eXperiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557) provides a unique laboratory for studying plasma instabilities and their relation to electron thermal transport due to its low toroidal field, high plasma beta, low aspect ratio and large ExB flow shear. Recent findings on NSTX have shown that multiple instabilities are required to explain observed electron thermal transport, given the wide range of equilibrium parameters due to different operational scenarios and radial regions in fusion plasmas. Here we review the recent progresses in understanding anomalous electron thermal transport in NSTX and focus on mechanisms that could drive electron thermal transport in the core region. The synergy between experiment and theoretical/ numerical modeling is essential to achieving these progresses. The plans for newly commissioned NSTX-Upgrade will also be discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2018

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 J 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 modelling 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 in this paper, 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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Stability and transport of gyrokinetic critical pedestals

Abstract A gyrokinetic threshold model for pedestal width–height scaling prediction is applied to multiple devices. A shaping and aspect ratio scan is performed on National Spherical Torus Experiment (NSTX) equilibria, finding Δ ped = 0.92 A 1.04 κ − 1.24 0.38 δ β θ , ped 1.05 for the wide-pedestal branch with pedestal width Δ ped , aspect ratio A , elongation κ , triangularity δ , and normalized pedestal height β θ , ped . The width–transport scaling is found to vary significantly if the pedestal height is varied either with a fixed density or fixed temperature, showing how fueling and heating sources affect the pedestal density and temperature profiles for the kinetic-ballooning-mode (KBM) limited profiles. For an NSTX equilibrium, at fixed density, the wide branch is Δ ped = 0.028 ( q e / Γ e − 1.7 ) 1.5 ∼ η e 1.5 and at fixed temperature Δ ped = 0.31 ( q e / Γ e − 4.7 ) 0.85 ∼ η e 0.85 , where q e and Γ e are turbulent electron heat and particle fluxes and η e = ∇ ln ⁡ T e / ∇ ln ⁡ n e for an electron temperature T e and density n e . Pedestals close to the KBM limit are shown to have modified turbulent transport coefficients compared to the strongly driven KBMs. The role of flow shear is studied as a width–height scaling constraint and pedestal saturation mechanism for a standard and lithiated wide pedestal discharge. Finally, the stability, transport, and flow shear constraints are combined and examined for an NSTX experiment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

An overview of recent physics results from NSTX

Currently, the National Spherical Torus Experiment (NSTX) is being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam (NB) for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-Upgrade to achieve the research goals critical to a Fusion Nuclear Science Facility. These include producing stable, 100% non-inductive operation in high-performance plasmas, assessing plasma-material interface (PMI) solutions to handle the high heat loads expected in the next-step devices and exploring the unique spherical torus (ST) parameter regimes to advance predictive capability. Non-inductive operation and current profile control in NSTX-U will be facilitated by co-axial helicity injection (CHI) as well as radio frequency (RF) and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet-Parker theory. Full-wave AORSA simulations show that RF power losses in the scrape-off layer (SOL) increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. Moreover, Toroidal Alfven eigenmode avalanches and higher frequency Alfven eigenmodes can affect NB-driven current through energy loss and redistribution of fast ions. The inclusion of rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing ideal wall and resistive wall modes. Neutral beams and neoclassical toroidal viscosity generated from applied 3D fields can be used as actuators to produce rotation profiles optimized for global stability. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade for disruption mitigation. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat flux handling. Simulations indicate that snowflake and impurity seeded radiative divertors are candidates for heat flux mitigation in NSTX-U. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium-oxygen affinity, which increases deuterium pumping and reduces recycling. In situ and test-stand experiments of lithiated graphite and molybdenum indicate temperature-enhanced sputtering, although that test-stand studies also show the potential for heat flux reduction through lithium vapour shielding. Non-linear gyro kinetic simulations have indicated that ion transport can be enhanced by a shear-flow instability, and that non-local effects are necessary to explain the observed rapid changes in plasma turbulence. Predictive simulations have shown agreement between a microtearing-based reduced transport model and the measured electron temperatures in a microtearing unstable regime. Finally, two Alfven eigenmode-driven fast ion transport models have been developed and successfully benchmarked against NSTX data. Upgrade construction is moving on schedule with initial physics research operation of NSTX-U planned for mid-2015.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Hydrogen Retention in Lithium on Metallic Walls from “In Vacuo” Analysis in LTX and Implications for High-Z Plasma-Facing Components in NSTX-U

The application of lithium to plasma-facing components (PFCs) has long been used as a technique for wall conditioning in magnetic confinement devices to improve plasma performance. Determining the characteristics of PFCs at the time of exposure to the plasma, however, is difficult because they can only be analyzed after venting the vacuum vessel and removing them at the end of an operational period. The Materials Analysis and Particle Probe (MAPP) addresses this problem by enabling PFC samples to be exposed to plasmas, and then withdrawn into an analysis chamber without breaking vacuum. The MAPP system was used to introduce samples that matched the metallic PFCs of the Lithium Tokamak Experiment (LTX). Lithium that was subsequently evaporated onto the walls also covered the MAPP samples, which were then subject to LTX discharges. In vacuo extraction and analysis of the samples indicated that lithium oxide formed on the PFCs, but improved plasma performance persisted in LTX. The reduced recycling this suggests is consistent with separate surface science experiments that demonstrated deuterium retention in the presence of lithium oxide films. Since oxygen decreases the thermal stability of the deuterium in the film, the release of deuterium was observed below the lithium deuteride dissociation temperature. This may explain what occurred when lithium was applied to the surface of the NSTX Liquid Lithium Divertor (LLD). The LLD had segments with individual heaters, and the deuterium-alpha emission was clearly lower in the cooler regions. The plan for NSTX-U is to replace the graphite tiles with high-Z PFCs, and apply lithium to their surfaces with lithium evaporation. Experiments with lithium coatings on such PFCs suggest that deuterium could still be retained if lithium compounds form, but limiting their surface temperatures may be necessary.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Hydrogen retention in lithium on metallic walls from “in vacuo” analysis in LTX and implications for high-Z plasma-facing components in NSTX-U

The application of lithium to plasma-facing components (PFCs) has long been used as a technique for wall conditioning in magnetic confinement devices to improve plasma performance. Determining the characteristics of PFCs at the time of exposure to the plasma, however, is difficult because they can only be analyzed after venting the vacuum vessel and removing them at the end of an operational period. The Materials Analysis and Particle Probe (MAPP) addresses this problem here by enabling PFC samples to be exposed to plasmas, and then withdrawn into an analysis chamber without breaking vacuum. The MAPP system was used to introduce samples that matched the metallic PFCs of the Lithium Tokamak Experiment (LTX). Lithium that was subsequently evaporated onto the walls also covered the MAPP samples, which were then subject to LTX discharges. In vacuo extraction and analysis of the samples indicated that lithium oxide formed on the PFCs, but improved plasma performance persisted in LTX. The reduced recycling this suggests is consistent with separate surface science experiments that demonstrated deuterium retention in the presence of lithium oxide films. Since oxygen decreases the thermal stability of the deuterium in the film, the release of deuterium was observed below the lithium deuteride dissociation temperature. This may explain what occurred when lithium was applied to the surface of the NSTX Liquid Lithium Divertor (LLD). The LLD had segments with individual heaters, and the deuterium-alpha emission was clearly lower in the cooler regions. The plan for NSTX-U is to replace the graphite tiles with high-Z PFCs, and apply lithium to their surfaces with lithium evaporation. Experiments with lithium coatings on such PFCs suggest that deuterium could still be retained if lithium compounds form, but limiting their surface temperatures may be necessary.

42 ENGINEERING↗

DOE OSTI2023

The effect of gas injection location on a lithium vapor box divertor in NSTX-U

Abstract The lithium vapor box divertor is a proposed divertor design that will minimize contamination of the upstream plasma in a fusion device, while also ensuring protection of the target. In this design lithium is evaporated near the target by high temperature lithium surfaces, dissipating the plasma heat flux. The lithium vapor box has been predicted via the fluid-kinetic code scrape off layer plasma simulator (SOLPS-ITER) to achieve low ( n L i / n e ∼ 0.05) upstream concentrations of lithium and low target heat fluxes. Here we compare two choices of deuterium gas puff location using SOLPS-ITER, the private flux region (PFR) and the common flux region (CFR), and find significant differences in the contamination level required to reach an acceptable target heat flux (defined here as q tar max ⩽ 10 MW m −2 ). Deuterium gas puffing from the PFR is seen to better reduce upstream lithium contamination. The difference in puffing location is seen to cause changes in the upstream flow of lithium ions. The PFR puff, having better access to the separatrix, can better reduce the upstream-directed flow of lithium near the separatrix which is the primary source of contamination due to a large thermal force in this region. Puffing from the CFR, partially due to inefficacy at reducing separatrix lithium flow, has higher lithium concentration within the plasma. Solutions that reduce the heat flux to below 10 MW m −2 have a range of lithium concentrations between n L i / n e ∼ 0.01–0.12 depending on puff intensity, location, evaporator temperature and recycling at the various plasma facing components. The efficacy of the puffs is tested for sensitivity to deuterium recycling coefficient at the target, evaporators, and main chamber walls. A CFR located puff is found to be more dependent on the recycling coefficients used than a PFR located puff. regardless of the set of recycling coefficients chosen, PFR puffing achieves lower lithium contamination than CFR puffing for a given heat flux.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. As a result, two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

NSTX-U Advances in Real-Time C++11 on Linux

Programming languages like C and Ada combined with proprietary embedded operating systems have dominated the real-time application space for decades. The new C++11standard includes native, language-level support for concurrency, a required feature for any nontrivial event-oriented real-time software. Threads, Locks, and Atomics now exist to provide the necessary tools to build the structures that make up the foundation of a complex real-time system. The National Spherical Torus Experiment Upgrade (NSTX-U) at the Princeton Plasma Physics Laboratory (PPPL) is breaking new ground with the language as applied to the needs of fusion devices. A new Digital Coil Protection System (DCPS) will serve as the main protection mechanism for the magnetic coils, and it is written entirely in C++11 running on Concurrent Computer Corporation's real-time operating system, RedHawk Linux. It runs over 600 algorithms in a 5 kHz control loop that determine whether or not to shut down operations before physical damage occurs. To accomplish this, NSTX-U engineers developed software tools that do not currently exist elsewhere, including real-time atomic synchronization, real-time containers, and a real-time logging framework. Together with a recent (and carefully configured) version of the GCC compiler, these tools enable data acquisition, processing, and output using a conventional operating system to meet a hard real-time deadline (that is, missing one periodic is a failure) of 200 microseconds.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

Erosion and re-deposition of lithium and boron coatings under high-flux plasma bombardment

Lithium and boron coatings are applied to the walls of many tokamaks to enhance performance and protect the underlying substrates. Li and B-coated high-Z substrates are planned for use in NSTX-U and are a candidate plasma-facing component (PFC) for DEMO. However, previous measurements of Li evaporation and thermal sputtering on low-flux devices indicate that the Li temperature permitted on such devices may be unacceptably low. Thus it is crucial to characterize gross and net Li erosion rates under high-flux plasma bombardment. Additionally, no quantitative measurements have been performed of the erosion rate of a boron-coated PFC during plasma bombardment. A realistic model for the compositional evolution of a Li layer under D bombardment was developed that incorporates adsorption, implantation, and diffusion. A model was developed for temperature-dependent mixed-material Li-D erosion that includes evaporation, physical sputtering, chemical sputtering, preferential sputtering, and thermal sputtering. The re-deposition fraction of a Li coating intersecting a linear plasma column was predicted using atomic physics information and by solving the Li continuity equation. These models were tested in the Magnum-PSI linear plasma device at ion fluxes of 10^23-10^24 m^-2 s^-1 and Li surface temperatures less than 800 degrees C. Li erosion was measured during bombardment with a neon plasma that will not chemically react with Li and the results agreed well with the erosion model. Next the ratio of the total D fluence to the areal density of the Li coating was varied to quantify differences in Li erosion under D plasma bombardment as a function of the D concentration. The ratio of D/Li atoms was calculated using the results of MD simulations and good agreement is observed between measurements and the predictions of the mixed-material erosion model. Li coatings are observed to disappear from graphite much faster than from TZM Mo, indicating that fast Li diffusion into the bulk graphite substrate occurred, as predicted. Li re-deposition fractions very close to unity are observed in Magnum-PSI, as predicted by modeling. Finally, predictions of Li coating lifetimes in the NSTX-U divertor are calculated. The gross erosion rate of boron coatings was also measured for the first time in a high-flux plasma device.

liquid metals↗

DOE OSTI2015

Electromagnetic Torque in Tokamaks with Toroidal Asymmetries

liquid metals↗

DOE OSTI2022

Application of simulations and theory of sub-cyclotron frequency modes to DIII-D

Theoretical and numerical study of Alfvén eigenmodes (AEs) in the sub-cyclotron frequency range explains the observed frequency scaling with beam parameters in experiments across different devices. Global Alfvén (GAEs) and compressional Alfvén eigenmodes (CAEs) are frequently excited during neutral beam injection (NBI) in the National Spherical Torus Experiment (NSTX/NSTX-U), as well as other beam-heated devices such as MAST and DIII-D. These modes are driven unstable through the Doppler shifted cyclotron resonance with the NBI ions and can be excited in ITER due to super-Alfvénic velocities and anisotropy of the fast ion distribution. Numerical models and theory for sub-cyclotron frequency modes, previously developed and used to study the excitation of GAEs/CAEs in the NSTX(-U), have been successfully applied to explain the DIII-D observations. Simulations for DIII-D demonstrate that the modes with ω/ω ci ≈ 0.6, previously mis-identified as compressional Alfvén eigenmodes, have shear polarization and should be identified as the GAEs. Simulation results match the observed frequencies and estimated toroidal mode numbers in DIII-D experiments. Good agreement is found between scaling of the experimentally observed GAE frequencies with NBI parameters for NSTX, NSTX-U and DIII-D and scaling predicted by the theory and simulations. Here, the DIII-D simulation results also demonstrate that GAEs can be excited for larger aspect-ratio devices and smaller (sub- Alfvénic) beam injection velocities, therefore they can be unstable in ITER.

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↗