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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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Elemental and topographical imaging of microscopic variations in deposition on NSTX-U and DIII-D samples2

Abstract: Tokamak plasma facing components have surface roughness that can cause microscopic spatial variations in erosion and deposition and hence influence material migration, erosion lifetime, dust and tritium accumulation, and plasma contamination. However high spatial resolution measurements of deposition on the scale of the surface roughness have been lacking to date. We will present elemental images of graphite samples from NSTX-U and DIII-D DiMES experiments performed with a Scanning Auger Microprobe at sub-micron resolution that show strong microscopic variations in deposition and correlate this with 3D topographical maps of surface irregularities. The NSTX-U samples were boronized and exposed to deuterium plasmas and the DiMES samples had localized Al and W films and were exposed to dedicated helium plasmas. Topographical maps of the samples were performed with a 3D confocal optical microscope and compared to the elemental deposition pattern. The results revealed localized deposition concentrated in areas shadowed from the ion flux, incident in a direction calculated (for the DiMES case) by taking account of the magnetic pre-sheath.

Erosion, Deposition, NSTX-U, DIMES, Plasma-Materia↗
DOE OSTI2019

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

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

current profile control↗
DOE OSTI2020

Toward fusion plasma scenario planning for NSTX-U using machine-learning-accelerated models

One of the most promising devices for realizing power production through nuclear fusion is the tokamak. To maximize performance, it is preferable that tokamak reactors achieve advanced operating scenarios characterized by good plasma confinement, improved magnetohydrodynamic (MHD) stability, and a largely non-inductively driven plasma current. Such scenarios could enable steady-state reactor operation with high \emph{fusion gain} --- the ratio of produced fusion power to the external power provided through the plasma boundary. Precise and robust control of the evolution of the plasma boundary shape as well as the spatial distribution of the plasma current, density, temperature, and rotation will be essential to achieving and maintaining such scenarios. The complexity of the evolution of tokamak plasmas, arising due to nonlinearities and coupling between various parameters, motivates the use of model-based control algorithms that can account for the system dynamics. In this work, a learning-based accelerated model trained on data from the National Spherical Torus Experiment Upgrade (NSTX-U) is employed to develop planning and control strategies for regulating the density and temperature profile evolution around desired trajectories. The proposed model combines empirical scaling laws developed across multiple devices with neural networks trained on empirical data from NSTX-U and a database of first-principles-based computationally intensive simulations. The reduced execution time of the accelerated model will enable practical application of optimization algorithms and reinforcement learning approaches for scenario planning and control development. An initial demonstration of applying optimization approaches to the learning-based model is presented, including a strategy for mitigating the effect of leaving the finite validity range of the accelerated model. The approach shows promise for actuator planning between experiments and in real-time.

machine learning↗
DOE OSTI2021

Dynamics of filaments during the edge-localized mode crash on NSTX

Edge localized modes (ELMs) are routinely observed in H-mode plasma regimes of the National Spherical Torus Experiment (NSTX). Due to the explosive nature of the instability, only diagnostics with high temporal and spatial resolution could provide a detailed insight into the dynamics associated with the ELMs. Gas-puff imaging (GPI) at NSTX provides 2D measurements of the magnetic field aligned fluctuations (e.g. ELM filaments) in the scrape-off layer and the at the plasma edge with 2.5 us temporal and 10 mm optical resolution.A novel analysis technique was developed to estimate the frame-by-frame velocities and the spatial parameters of the dominant structures associated with the ELMs. The analysis was applied to single ELM events to characterize the ELM crash dynamics, and then extended to a database of 169 ELM events.Statistical analysis was performed in order to find the characterizing dynamics of the ELM crash. The results show that on average an ELM crash consists of a filament with a circular cross-section which is propelled outwards with a characterizing peak radial velocity of ~3.3 km/s. The radial velocity was found to be linearly dependent on the distance of the filament from the separatrix, which has never been seen before. The ELM filament is characterized by propagation in the ion-diamagnetic direction poloidally with a peak velocity of 11.4 km/s. The ELM crash lasts for approximately 100us until the radial propulsion settles back to the pre-ELM level. The experimental findings were compared with analytical theory. Two possible mechanisms were identified for explaining the observations: the curvature interchange model and the current-filament interaction model.

correlation↗
DOE OSTI2012

Comparison of Gas Puff Imaging Data in NSTX with the DEGAS 2 Simulation

Gas-Puff-Imaging (GPI) is a two dimensional diagnostic which measures the edge D α light emission from a neutral D 2 gas puff nears the outer mid-plane of NSTX. DEGAS 2 is a 3-D Monte Carlo code used to model neutral transport and atomic physics in tokamak plasmas. In this paper we compare measurements of the D α light emission obtained by GPI on NSTX with DEGAS 2 simulations of D α light emission for specific experiments. Both the simulated spatial distribution and absolute intensity of the D α light emission agree well with the experimental data obtained between ELMs in H-mode.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Detection of Disruptions in the High-β Spherical Torus NSTX

This paper describes the prediction of disruptions based on diagnostic data in the high-β spherical torus NSTX. The disruptive threshold values on many signals are examined. In some cases, raw diagnostic data can be used as a signal for disruption prediction. In others, the deviations of the plasma data from simple models provides the signal used to determine the proximity to disruption. However, no single signal and threshold value can form the basis for disruption prediction in NSTX; thresholds that produce an acceptable false positive rate have too large a missed or late warning rate, while combinations that produce an acceptable rate of missed or late warnings have an unacceptable false positive rate. To solve this problem, a novel means of combining multiple threshold tests has been developed. After being properly tuned, this algorithm can produce a false positive rate of 2.8%, with a late warning rate of 3.7% when applied to a database of ~2000 disruptions collected from three run campaigns. Furthermore, many of these false positives are triggered by near-disruptive MHD events that might indeed be disruptive in larger plasmas with more stored energy. However, the algorithm is less efficient at detecting the MHD event that prompts the disruption process.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Progress in Simulating Turbulent Electron Thermal Transport in NSTX

Nonlinear simulations based on multiple NSTX discharge scenarios have progressed to help differentiate unique instability mechanisms and to validate with experimental turbulence and transport data. First nonlinear gyrokinetic simulations of microtearing (MT) turbulence in a high-beta NSTX H-mode discharge predict experimental levels of electron thermal transport that are dominated by magnetic flutter and increase with collisionality, roughly consistent with energy confinement times in dimensionless collisionality scaling experiments. Electron temperature gradient (ETG) simulations predict significant electron thermal transport in some low and high beta discharges when ion scales are suppressed by E x B shear. Although the predicted transport in H-modes is insensitive to variation in collisionality (inconsistent with confinement scaling), it is sensitive to variations in other parameters, particularly density gradient stabilization. In reversed shear (RS) Lmode discharges that exhibit electron internal transport barriers, ETG transport has also been shown to be suppressed nonlinearly by strong negative magnetic shear, s<<0. In many high beta plasmas, instabilities which exhibit a stiff beta dependence characteristic of kinetic ballooning modes (KBM) are sometimes found in the core region. However, they do not have a distinct finite beta threshold, instead transitioning gradually to a trapped electron mode (TEM) as beta is reduced to zero. Nonlinear simulations of this "hybrid" TEM/KBM predict significant transport in all channels, with substantial contributions from compressional magnetic perturbations. As multiple instabilities are often unstable simultaneously in the same plasma discharge, even on the same flux surface, unique parametric dependencies are discussed which may be useful for distinguishing the different mechanisms experimentally.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Fast-wave Power Flow Along SOL Field Lines In NSTX nd The Associated Power Deposition Profile Across The SOL In Front Of The Antenna

Fast-wave heating and current drive efficiencies can be reduced by a number of processes in the vicinity of the antenna and in the scrape off layer (SOL). On NSTX from around 25% to more than 60% of the high-harmonic fast-wave power can be lost to the SOL regions, and a large part of this lost power flows along SOL magnetic field lines and is deposited in bright spirals on the divertor floor and ceiling. We show that field-line mapping matches the location of heat deposition on the lower divertor, albeit with a portion of the heat outside of the predictions. The field-line mapping can then be used to partially reconstruct the profile of lost fast-wave power at the midplane in front of the antenna, and the losses peak close to the last closed flux surface (LCFS) as well as the antenna. This profile suggests a radial standing-wave pattern formed by fast-wave propagation in the SOL, and this hypothesis will be tested on NSTX-U. Advanced RF codes must reproduce these results so that such codes can be used to understand this edge loss and to minimize RF heat deposition and erosion in the divertor region on ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Using LGI experiments to achieve better understanding of pedestal-edge coupling in NSTX-U

PowerPoint presentation. Latest advances in granule or dust injection technologies, fast and high-resolution imaging, together with micro-/nano-structured material fabrication, provide new opportunities to examine plasma-material interaction (PMI) in magnetic fusion environment. Some of our previous work in these areas is summarized. The upcoming LGI experiments in NSTX-U will shed new light on granular matter transport in the pedestal-edge region. In addition to particle control, these results can also be used for code validation and achieving better understanding of pedestal-edge coupling in fusion plasmas in both NSTX-U and others.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

The Motional Stark Effect Diagnostic for NSTX-U

This report describes the development of the motional Stark Effect diagnostic for NSTX. The MSE concept relies upon observation of the Balmer-alpha (n=3 to n=2) emission from a neutral hydrogen beam traversing a plasma. As a beam at high velocity, $\vec{ν}$, passes through the magnetic field in the plasma, $\vec{B}$, it experiences in its reference frame a Lorentz electric field, $\vec{E}$ = $\vec{ν}$ x $\vec{B}$. This electric field causes the spectral emission to be split and polarized as described by the Stark effect. The transition is separated into a set of lines in which the line spacing is linearly proportional to the electric field magnitude, and the line polarization is dependent on the electric field direction. Traditionally, MSE has used polarimetry to determine the polarization angle of the light, and related that to the magnetic field pitch angle in the plasma. The MSE diagnostic was developed under high magnetic field ($>$ 1~T) conditions, and the lower fields of recent high-beta experiments such as NSTX posed challenges to the technique. At low magnetic fields Doppler broadening effects mask the Stark effect. With the use of a narrow bandwidth filter MSE measurements at very low magnetic fields have been obtained.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Easy Web Interfaces to IDL Code for NSTX Data Analysis

Reusing code is a well-known Software Engineering practice to substantially increase the efficiency of code production, as well as to reduce errors and debugging time. A variety of "Web Tools" for the analysis and display of raw and analyzed physics data are in use on NSTX [1], and new ones can be produced quickly from existing IDL [2] code. A Web Tool with only a few inputs, and which calls an IDL routine written in the proper style, can be created in less than an hour; more typical Web Tools with dozens of inputs, and the need for some adaptation of existing IDL code, can be working in a day or so. Efficiency is also increased for users of Web Tools because o f the familiar interface of the web browser, and not needing X-windows, accounts, passwords, etc. Web Tools were adapted for use by PPPL physicists accessing EAST data stored in MDSplus with only a few man-weeks of effort; adapting to additional sites should now be even easier. An overview of Web Tools in use on NSTX, and a list of the most useful features, is also presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Predictions of VRF on a Langmuir Probe under the RF Heating Spiral on the Divertor Floor on NSTX-U

RF heating deposition spirals are observed on the divertor plates on NSTX as shown in for a NB plus RF heating case. It has been shown that the RF spiral is tracked quite well by the spiral mapping of the strike points on the divertor plate of magnetic field lines passing in front of the high harmonic fast wave (HHFW) antenna on NSTX. Indeed, both current instrumented tiles and Langmuir probes respond to the spiral when it is positioned over them. In particular, a positive increment in tile current (collection of electrons) is obtained when the spiral is over the tile. This current can be due to RF rectification and/or RF heating of the scrape off layer (SOL) plasma along the magnetic field lines passing in front of the the HHFW antenna. It is important to determine quantitatively the relative contributions of these processes. Here we explore the properties of the characteristics of probes on the lower divertor plate to determine the likelyhood that the primary cause of the RF heat deposition is RF rectification.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

The Role of Lithium Conditioning in Achieving High Performance, Long Pulse H-mode Discharges in the NSTX and EAST Devices

In this paper, the role of lithium wall conditioning on the achievement of high performance, long pulse discharges in the National Spherical Torus Experiment (NSTX) and the Experimental Advanced Superconducting Tokamak (EAST) is documented. Common observations include recycling reduction and elimination of ELMs. In NSTX, lithium conditioning typically resulted in ELM-free operation with impurity accumulation, which was ameliorated e.g. with pulsed 3D fields to trigger controlled ELMs. Active lithium conditioning in EAST discharges has overcome this problem, producing an ELM-free Hmode with controlled density and impurities.

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 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 OSTI2018

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

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

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

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

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

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Enhancing operational safety: The NSTX-U Shorted Turn Protection (STP) system

The NSTX-U Shorted Turn Protection (STP) system is a vital safety feature designed to protect the tokamak’s coils during operations. It employs a Kalman filter-based algorithm to monitor coil currents and detect faults in real-time. If an anomaly is detected, the system swiftly terminates the tokamak pulse to prevent damage. Developed with Matlab and Simulink, the STP system has been validated through real-time simulations, ensuring its effectiveness for the upcoming upgrades of the NSTX-U facility. In conclusion, this article presents the theory for implementation of the STP algorithm and the results from simulation and real-time implementation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗