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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 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Visualization of Fast Ion Phase-Space Flow Driven by Alfvén Instabilities

Fast ion phase-space flow, driven by Alfven eigenmodes (AEs), is measured by an imaging neutral particle analyzer in the DIII-D tokamak. The flow firstly appears near the minimum safety factor at the injection energy of neutral beams, and then moves radially inward and outward by gaining and losing energy, respectively. The flow trajectories in phase space align well with the intersection lines of the constant magnetic moment surfaces and constant E – (ω/n)P ζ surfaces, where E, P ζ are energy and toroidal canonical momentum of ions; ω and n are angular frequencies and toroidal mode numbers of AEs. It is found that the flow is so destructive that the thermalization of fast ions is no longer observed in regions of strong interaction. Here, the measured phase-space flow is consistent with nonlinear hybrid kinetic-magnetohydrodynamics simulation. Calculations of the relatively narrow phase-space islands reveal that fast ions must transition between different flow trajectories to experience large-scale phase-space transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Experimental Validation of a Kinetic Ballooning Mode in High-Performance High-Bootstrap Current Fraction Fusion Plasmas

We report the observation of a set of coherent high frequency electromagnetic fluctuations that leads to a turbulence induced self-regulating phenomenon in the DIII-D high bootstrap current fraction plasma. The fluctuations have frequency of 130~220kHz, the poloidal wave length and phase velocity are 16~30 m -1 and ~30 km/s, respectively in the outboard midplane with the estimated toroidal mode number n~5- 9. The fluctuations are located in the internal transport barrier (ITB) region at large radius and are experimentally validated to be kinetic ballooning modes (KBM). Furthermore, quasilinear estimation predicts the KBM to be able to drive experimental particle flux and non-negligible thermal flux, suggesting its significant role in regulating the ITB saturation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Tokamak Plasmas with Density up to 10 Times the Greenwald Limit

Current-carrying, toroidal laboratory plasmas typically cannot be sustained with an electron density above the empirical Greenwald limit. Presented here are tokamak experiments in the Madison Symmetric Torus with a density up to an unprecedented level about ten times this limit. This is thought to be made possible in part by a thick, stabilizing, conductive wall, and a high-voltage, feedback-controlled power supply driving the plasma current. The radial profile of the toroidal current flattens around twice the limit, without the edge collapse routinely observed in other experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Efficiency and Physical Limitations of Adiabatic Direct Energy Conversion in Axisymmetric Fields

We describe and analyze a new class of direct energy conversion schemes based on the adiabatic magnetic drift of charged particles in axisymmetric magnetic fields. The efficiency of conversion as well as the geometrical and dynamical limitations of the recoverable power are calculated. The geometries of these axisymmetric field configurations are suited for direct energy conversion in radiating advanced aneutronic reactors and in advanced divertors of deuterium-tritium tokamak reactors. The E × B configurations considered here do not suffer from the classical drawbacks and limitations of thermionic and magnetohydrodynamic high-temperature direct energy conversion devices.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2021

Real-Time and Adaptive Reservoir Computing With Application to Profile Prediction in Fusion Plasma

Nuclear fusion is a promising alternative to address the problem of sustainable energy production. The tokamak is an approach to fusion based on magnetic plasma confinement, constituting a complex physical system with many control challenges. Here we study the characteristics and optimization of reservoir computing (RC) for real-time and adaptive prediction of plasma profiles in the DIII-D tokamak. Our experiments demonstrate that RC achieves comparable results to state-of-the-art (deep) convolutional neural networks (CNNs) and long short-term memory (LSTM) models, with a significantly easier and faster training procedure. This efficient approach allows for fast and frequent adaptation of the model to new situations, such as changing plasma conditions or different fusion devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Developments and Challenges in the Design of the ITER DRGA

We report the ITER Diagnostic Residual Gas Analyzer (DRGA) will measure the distribution of gas species, i.e., deuterium (D), tritium (T), and impurities, in the divertor exhaust stream and in the plasma periphery, with time resolution relevant to fusion plasma–wall particle dynamics. The uniqueness of the DRGA, over previous implementations of plasma dynamics residual gas analysis, is an integrated approach, combining mass and low-temperature plasma-activated optical spectroscopy, in a differentially pumped analysis station. A further unique feature of the ITER divertor-specific DRGA is an ~8-m separation of the analysis station from the sampled pumping duct, while still maintaining a ~1-s response time for hydrogen isotopic concentrations. ITER DRGA final design activities are strongly benefiting from testing of prototypical DRGA components and methods on present fusion devices, most currently on JET and W7-X. DRGA systems are implemented on both these devices and include sensors (and pumping methods) that are directly relevant to the ITER DRGA design. The recent JET-DTE2 campaign has provided the first experience on operating the combined ITER DRGA sensors with D-T plasmas. While enhancing system design for ITER, this experience on operating devices has also revealed additional engineering challenges, which further guide the continuing final design project. Meanwhile, the recent determination that the ITER DRGA, with slight optimization, will resolve the helium isotopes well enough to support an ITER pre-DT, He-3-based heating scheme, has greatly increased ITER Research Program interest in the DRGA and its implementation well ahead of the DT phase.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

A software package for plasma facing component analysis and design: the Heat flux Engineering Analysis Toolkit (HEAT)

The engineering limits of plasma facing components (PFCs) constrain the allowable operational space of tokamaks. Poorly managed heat fluxes that push the PFCs beyond their limits not only degrade core plasma performance via elevated impurities, but can also result in PFC failure due to thermal stresses or melting. Simple axisymmetric assumptions fail to capture the complex interaction between 3D PFC geometry and 2D or 3D plasmas. This results in fusion systems that must either operate with increased risk or reduce PFC loads, potentially through lower core plasma performance, to maintain a nominal safety factor. High precision 3D heat flux predictions are necessary to accurately ascertain the state of a PFC given the evolution of the magnetic equilibrium. A new code, the Heat flux Engineering Analysis Toolkit (HEAT), has been developed to provide high precision 3D predictions and analysis for PFCs. HEAT couples many otherwise disparate computational tools together into a single open source python package. Magnetic equilibrium, engineering CAD, finite volume solvers, scrape off layer plasma physics, visualization, high performace computing, and more, are connected in a single web-based user interface. Linux users may use HEAT without any software prerequisites via an appImage. This manuscript introduces HEAT, discusses the software architecture, presents first HEAT results, and outlines physics modules in development.

divertor physics↗
DOE OSTI2016

AE3D

AE3D solves for the shear Alfven eigenmodes and eigenfrequencies in a torodal magnetic fusion confinement device. The configuration can be either 2D (e.g. tokamak, reversed field pinch) or 3D (e.g. stellarator, helical reversed field pinch, tokamak with ripple). The equations solved are based on a reduced MHD model and sound wave coupling effects are not currently included.

Spong, DonaldA↗
DOE OSTI2016

Close-out report with links to abstracts

This grant provided A/V support for two technical meetings of the Edge Coordinating Committee: (1) Nov 13, 2013 (co-located with the APS-DPP meeting in Denver, CO) https://ecc.mit.edu/fall-2013-technical-meeting#overlay-context=ecc-meetings; (2) April 28-May 1, 2015 (embedded sessions in the Transport Task Force Meeting, Salem, MA) http://www-internal.psfc.mit.edu/TTF2015/index.html. The ultimate goal of the U.S. Transport Task Force is to develop a physics-based understanding of particle, momentum and heat transport in magnetic fusion devices. This understanding should be of sufficient depth that it allows the development of predictive models of plasma transport that can be validated against experiment, and then used to anticipate the future performance of burning plasmas in ITER, as well as to provide guidance for the design of next-step fusion nuclear science facilities. To achieve success in transport science, it is essential to characterize local fluctuations and transport in toroidal plasmas, to understand the basic mechanisms responsible for transport, and ultimately to control these transport processes. These goals must be pursued in multiple areas, and these topics evolve in order to reflect current interests.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Alcator C-Mod (Final Technical Report)

Alcator C-Mod is the third in a series of compact high-field tokamaks built and operated on the MIT campus, the first to incorporate advanced shaping and poloidal divertors. For most of its 23 year operational history, C-Mod was the only diverted tokamak in the world program exclusively using all metal high-Z plasma facing components, and the only diverted tokamak able to operate at and above the magnetic field strength of ITER, and for many other burning plasma tokamak designs. The C-Mod facility also exclusively used RF waves for non-inductive heating and current drive, highly relevant to burning plasma applications and likely to be the main such tools in any commercial tokamak power reactor. A comprehensive summary of the key technical accomplishments from the C-Mod program is published in the review paper, M. Greenwald, et al., Phys. Plasmas 21(2014)110501. This final report contains a brief overview of those results, along with updates covering the final 3 years of operation, through September 2016. Ongoing analysis and modeling based on the archive of Alcator C-Mod data continues to provide new results, and this can be expected to continue for many years to come.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Nonlinear and 3D plasmas. Final report

The project objectives are to characterize the nonlinear properties of edge instabilities in fusion plasmas from the standpoint of MHD and extended MHD models of plasmas and to develop, as much as possible, simplified models that may be useful for prediction and control.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Neutral Atomic Beam Techniques for Diagnosis of Edge and Scrape-Off Layer Plasmas (2017-2019) [and subsequently titled] Neutral Atomic Beam Techniques for Diagnosis of Interior and Scrape-Off Layer Plasmas (2019-2022) (Final Report)

This is the Final Technical Report for the DOE Measurement Innovation grant DE-SC0017998 titled Neutral Atomic Beam Techniques for Diagnosis of Edge and Scrape-Off Layer Plasmas (2017-2019) and subsequently titled Neutral Atomic Beam Techniques for Diagnosis of Interior and Scrape-Off Layer Plasmas (2019-2022). The grant was funded by the Office of Fusion Energy Sciences (FES) for the period 1 September 2017 – 31 May 2022. Work performed through the grant has resulted in diagnostic innovations and new measurement capabilities that could improve understanding of magnetic equilibrium, edge localized modes in tokamaks, and bootstrap and Pfirsch-Schlüter currents in stellarators. We have performed simulations of a novel non-hydrogen beam-based diagnostic that would determine a local value (in the plasma) of the poloidal flux function by measuring the toroidal angle of the beam particles' velocity at a detector external to the plasma. The approach is similar to that of a Heavy Ion Beam Probe (HIBP), but would be able to operate with a smaller, more economical detection systems and, in doing so, may enable substantial extension of scenarios in which beam-based diagnostics are deployed. Key results include: Identification of the neutral and singly charged atoms that are most appropriate for use in the beam based diagnostic and the ionization cross-sections that are important for studying the plasma edge of a large device; Development of a promising geometry for operation in which the beam is injected horizontally near the torus midplane; Implementation of simulations using banana orbit trajectories of secondary beam ions to access plasma regions that are commonly in-accessible to an HIBP diagnostic, including sample volumes near the X-point of a diverted tokamak; Assessment of the ability of the diagnostic concept (originally developed under the assumption of, and reliant on, symmetry) to work in stellarators by performing simulations with both quasi-symmetric fields and non-quasi-symmetric fields; Measurement of noise levels with a prototype detector on a stellarator and exploration of techniques to minimize effects of noise on a diagnostic.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Morphological Properties of the Crab Nebula: A Detailed Multiwavelength Study Based on New VLA, HST, Chandra, and XMM-Newton Images

We present a detailed analysis of the morphological properties of the Crab Nebula across the electromagnetic spectrum based on new and previous high-quality data from radio to X-rays. In the radio range we obtained an image of the entire nebula at 3 GHz with subarcsecond angular resolution using the VLA (NRAO) and an image at 100 GHz of the central region using the ALMA array. Simultaneously with the VLA observations we performed HST WFPC3 near-infrared (λ∼1.5 μm) and Chandra X-ray (0.5–8 keV band) observations of the central region of the nebula. In addition we produced a new UV image of the Crab Nebula at 291 nm by co-adding 75 individual exposures of the Optical-UV Monitor on board XMM-Newton. The high-angular resolution and high-dynamic range radio image at 3 GHz allowed us to improve the detection and characterization of peculiar morphological features including arches with foot brightening and intercrossed loop-like structures, likely originating in plasma confined to magnetic field lines. Based on the new radio image, we carried out a detailed multiwavelength correlation. In the central area, the comparison of the almost simultaneous images confirms that the wisps in the three spectral ranges do not generally coincide in location, the radio emission being the most discordant, which is suggestive of the existence of two different synchrotron components. The X-ray pulsar jet does not have a radio counterpart. Instead, another jet-like feature is seen in radio, though with different curvature and starting point.

79 ASTRONOMY AND ASTROPHYSICS↗