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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Full wave simulations of fast wave efficiency and power losses in the scrape-off layer of tokamak plasmas in mid/high harmonic and minority heating regimes

Here, several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves (HHFW), have found strong interaction between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 2D and 3D AORSA results for the National Spherical Torus eXperiment (NSTX) have shown a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is removed from in front of the antenna by increasing the edge density. Here, full wave simulations have been extended for 'conventional' tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results in HHFW regime show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for C-Mod and EAST, which operate in the minority heating regime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

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. Finally, 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.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2018

A Hybridizable Discontinuous Galerkin solver for the Grad–Shafranov equation

In axisymmetric fusion reactors, the equilibrium magnetic configuration can be expressed in terms of the solution to a semi-linear elliptic equation known as the Grad–Shafranov equation, the solution of which determines the poloidal component of the magnetic field. When the geometry of the confinement region is known, the problem becomes an interior Dirichlet boundary value problem. We propose a high order solver based on the Hybridizable Discontinuous Galerkin method. The resulting algorithm (1) provides high order of convergence for the flux function and its gradient, (2) incorporates a novel method for handling piecewise smooth geometries by extension from polygonal meshes, (3) can handle geometries with non-smooth boundaries and x-points, and (4) deals with the semi-linearity through an accelerated two-grid fixed-point iteration. The effectiveness of the algorithm is verified with computations for cases where analytic solutions are known on configurations similar to those of actual devices (ITER with single null and double null divertor, NSTX, ASDEX upgrade, and Field Reversed Configurations).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Full wave simulations of fast wave efficiency and power losses in the scrape-off layer of tokamak plasmas in mid/high harmonic and minority heating regimes

Several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves, have found strong interaction between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 3D AORSA results for National Spherical Torus eXperiment (NSTX), where a full antenna spectrum is reconstructed, are shown confirming the same behavior found for a single toroidal mode results in Bertelli et al Nucl. Fusion, 54 083004, 2014, namely, a strong transition to higher SOL power losses (driven by the RF field) has been found when the FW cut-off is removed from in front of the antenna by increasing the edge density. Additionally, full wave simulations have been extended for “conventional” tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for C-Mod and EAST, which operate in the minority heating regime unlike NSTX/NSTX-U and DIII-D, which operate in the mid/high harmonic regime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment

This thesis presents new measurements of core impurity concentrations and transport in plasmas with lithium coatings on all-metal plasma facing components (PFCs) in the Lithium Tokamak Experiment (LTX). LTX is a modest-sized spherical tokamak uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma (as opposed to just the divertor or limiter region in other devices). Lithium (Li) wall-coatings have improved plasma performance and confinement in several tokamaks with carbon (C) PFCs, including the National Spherical Torus Experiment (NSTX). In NSTX, contamination of the core plasma with Li impurities was very low (<0.1%) despite extensive divertor coatings. Low Li levels in NSTX were found to be largely due to neoclassical forces from the high level of C impurities. Studying impurity levels and transport with Li coatings on stainless steel surfaces in LTX is relevant to future devices (including future enhancements to NSTX-Upgrade) with all-metal PFCs. The new measurements in this thesis were enabled by a refurbished Thomson scattering system and improved impurity spectroscopy, primarily using a novel visible spectrometer monitoring several Li, C, and oxygen (O) emission lines. A simple model was used to account for impurities in unmeasured charge states, assuming constant density in the plasma core and constant concentration in the edge. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with~2-4% Li, ~0.6-2% C, ~0.4-0.7% O, and Z_eff<1.2. Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, unlike in NSTX, where collisions with C dominated. Furthermore, neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two, in contrast to NSTX where they differed by an order of magnitude. However, time-independent simulations with MIST indicated that unlike NSTX, neoclassical theory did not fully capture the impurity transport and anomalous transport likely played a significant role in determining impurity profiles.

Lithium plasma facing components↗
DOE OSTI2016

Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution]

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower n e , and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2 -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2 -seeded SF divertor at P SOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected P SOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider n e operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (cf. standard divertor).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2 -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2 -seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (see standard divertor).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Development of miniaturized, spectroscopically assisted Penning gauges for fractional helium and hydrogen neutral pressure measurements

Direct measurements of the helium (He) fractional neutral pressure in the neutral gas around fusion devices is challenging because of the small mass difference between the abundant D2 molecules and the He atoms which will be produced by D-T fusion. To study He exhaust, Penning gauges were used to measure total neutral pressure assisted by spectroscopy to resolve the D and He partial pressures. In this contribution, initial results are shown from developing this technique into a miniaturized configuration for direct in-situ measurements in the divertor of fusion devices. The configuration is based off a gauge originally designed for the National Spherical Tokamak Experiment-Upgrade (NSTX-U). The goal of this new miniaturized design it to reduce the space required by the gauge on the device and use of the inherent magnetic field of the machine rather than permanent magnets inside the gauge, enabling it to be adapted into a system that can be extended directly into the divertor region. The feasibility test of the method for NSTX-U and the Wendelstein 7-X (W7-X) stellarator are surveyed. For W7-X, a commercial Penning Gauge has been installed on an outboard vacuum flange as a generic feasibility test in the neutral gas environment of a stellarator. At an integration time of 25s, helium lines can be seen down to 10^-5 mbar and H-alpha lines down to 10^-6 mbar. Successful measurement of the total as well as the fractional neutral pressures of He and H has been shown. A first prototype of the miniature Penning gauge has been tested in Madison and shows a near linear power law scaling between current and pressure: I = C*P^n with n = 1.0 - 1.2. Pressure measurements were achieved starting at 10^-3 mbar and down to 10^-6 mbar. A modular gauge is being assembled, which allows easy interchangeability of the anode to test new anode geometries, in order to improve optical access and increase spectroscopic sensitivity. This shall enable an increase of the time resolution of the spectroscopically assisted fractional neutral pressure measurements to up to 1kHz.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Reduced model for direct induction startup scenario development on MAST-U and NSTX-U

A reduced semi-empirical model using time-dependent axisymmetric vacuum field calculations is used to develop the prefill and feed-forward coil current targets required for reliable direct induction (DI) startup on the new MA-class spherical tokamaks, MAST-U and NSTX-U. The calculations are constrained by operational limits unique to each device, such as the geometry of the conductive elements and active coils, power supply specifications and coil heating and stress limits. The calculations are also constrained by semi-empirical models for sufficient breakdown, current drive, equilibrium and stability of the plasma developed from a shared database. A large database of DI startup on NSTX and NSTX-U is leveraged to quantify the requirements for achieving a reliable breakdown (I p ~ 20 kA). It is observed that without pre-ionization, STs access the large E/P regime at modest loop voltage (V loop ) where the electrons in the weakly ionized plasma are continually accelerating along the open field lines. This ensures a rapid (order millisecond) breakdown of the neutral gas, even without pre-ionization or high-quality field nulls. The timescale of the initial increase in I p on NSTX is reproduced in the reduced model provided a mechanism for impeding the applied electric field is included. Most discharges that fail in the startup phase are due to an inconsistency in the evolution of the plasma current (I p ) and equilibrium field or loss of vertical stability during the burn-through phase. Here, the requirements for the self-consistent evolution of the fields in the weakly and fully-ionized plasma states are derived from demonstrated DI startup on NSTX, NSTX-U and MAST. The predictive calculations completed for MAST-U and NSTX-U illustrate that the maximum I p ramp rate (dI p /dt) in the early startup phase is limited by the voltage limits on the poloidal field coils on MAST-U and passive vertical stability on NSTX-U.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Reduced Model for Direct Induction Startup Scenario Development on MAST-U and NSTX-U

A reduced semi-empirical model using time-dependent axisymmetric vacuum field calculations is used to develop the prefill and feed-forward coil current targets required for reliable direct induction (DI) startup on the new MA-class spherical tokamaks, MAST-U and NSTX-U. The calculations are constrained by operational limits unique to each device, such as the geometry of the conductive elements and active coils, power supply specifications and coil heating and stress limits. The calculations are also constrained by semi-empirical models for sufficient breakdown, current drive, equilibrium and stability of the plasma developed from a shared database. A large database of DI startup on NSTX and NSTX-U is leveraged to quantify the requirements for achieving a reliable breakdown (Ip ~ 20 kA). It is observed that without pre-ionization, STs access the large E/P regime at modest loop voltage (Vloop) where the electrons in the weakly ionized plasma are continually accelerating along the open field lines. This ensures a rapid (order millisecond) breakdown of the neutral gas, even without pre-ionization or high-quality field nulls. The timescale of the initial increase in Ip on NSTX is reproduced in the reduced model provided a mechanism for impeding the applied electric field is included. Most discharges that fail in the startup phase are due to an inconsistency in the evolution of the plasma current (Ip) and equilibrium field or loss of vertical stability during the burn-through phase. The requirements for the self-consistent evolution of the fields in the weakly and full-ionized plasma states are derived from demonstrated DI startup on NSTX, NSTX-U and MAST. The predictive calculations completed for MAST-U and NSTX-U illustrate that the maximum Ip ramp rate (dIp/dt) in the early startup phase is limited by the voltage limits on the poloidal field coils on MAST-U and passive vertical stability on NSTX-U.

Direct induction↗
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↗