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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Plasma Facing Component Characterization and Correlation With Plasma Conditions in Lithium Tokamak Experiment-β

Lithium coatings in the Lithium Tokamak eXperiment (LTX) led to flat temperature profiles. The flat temperature profiles were observed along with a hot, low density edge, implying a broad, collisionless scrape-off layer (SOL). Additionally, in vacuo X-ray photoelectron spectroscopy (XPS) measurements established that lithium coatings evaporatively deposited onto high-Z plasma facing components (PFCs) became oxidized while retaining the ability to achieve good plasma performance long after lithium was applied to the PFCs. Longstanding theory predicted flat temperature profiles with low recycling walls, which was presumed to be due to hydrogen binding with elemental lithium to form lithium hydride. The presence of oxidized lithium, however, raised questions regarding the exact mechanism of hydrogen retention in LTX. To investigate these questions, the upgraded facility LTX- β includes a new sample exposure probe (SEP) for more detailed in vacuo analysis of PFC samples. The SEP is equipped with a vacuum suitcase capable of transporting samples representative of the LTX- β outer midplane PFCs to a stand-alone XPS system while maintaining pressures lower than the LTX- β base vacuum to limit the contamination between sample exposure and analysis. The low-energy resolution XPS system used in past experiments could only enable the determination of elemental percentages on the PFC sample surfaces. Because the new XPS system has higher energy resolution, it is more direct to assign chemical compounds to the measured binding energies. This capability has been confirmed by comparing XPS data from PFC test samples with measurements using a commercial high-resolution XPS system. Quartz crystal microbalances (QCMs) were used to quantify the thickness of the deposited lithium on the LTX- β PFCs. This article describes the application of the SEP to characterize the PFC surfaces using XPS and their relationship to plasma conditions.

Maan, Anurag↗
DOE OSTI2020

Projections of H-mode access and edge pedestal in the SPARC tokamak

In order to inform core performance projections and divertor design, the baseline SPARC tokamak plasma discharge is evaluated for its expected H-mode access, pedestal pressure and edge-localized mode (ELM) characteristics. A clear window for H-mode access is predicted for full field DT plasmas, with the available 25 MW of design auxiliary power. Additional alpha heating is likely needed for H-mode sustainment. Pressure pedestal predictions in the developed H-mode are surveyed using the EPED model. The projected SPARC pedestal would be limited dominantly by peeling modes and may achieve pressures in excess of 0.3 MPa at a density of approximately 3 × 10 20 m -3 . High pedestal pressure is partially enabled by strong equilibrium shaping, which has been increased as part of recent design iterations. Edge-localized modes (ELMs) with >1 MJ of energy are projected, and approaches for reducing the ELM size, and thus the peak energy fluence to divertor surfaces, are under consideration. The high pedestal predicted for SPARC provides ample margin to satisfy its high fusion gain (Q) mission, so that even if ELM mitigation techniques result in a 2x reduction of the pedestal pressure, Q > 2 is still predicted.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Synergistic cross-scale coupling of turbulence in a tokamak plasma

For the first time, nonlinear gyrokinetic simulations spanning both the ion and electron spatio-temporal scales have been performed with realistic electron mass ratio ((m D /m e ) 1/2 = 60.0), realistic geometry, and all experimental inputs, demonstrating the coexistence and synergy of ion ( ( k θ ρ s ∼ O ( 1.0 ) ) ) and electron-scale ( ( k θ ρ e ∼ O ( 1.0 ) ) ) turbulence in the core of a tokamak plasma. All multi-scale simulations utilized the GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] to study the coupling of ion and electron-scale turbulence in the core (r/a = 0.6) of an Alcator C-Mod L-mode discharge shown previously to exhibit an under-prediction of the electron heat flux when using simulations only including ion-scale turbulence. Electron-scale turbulence is found to play a dominant role in setting the electron heat flux level and radially elongated (k r ≪ k θ ) “streamers” are found to coexist with ion-scale eddies in experimental plasma conditions. Inclusion of electron-scale turbulence in these simulations is found to increase both ion and electron heat flux levels by enhancing the transport at the ion-scale while also driving electron heat flux at sub-ρ i scales. The combined increases in the low and high-k driven electron heat flux may explain previously observed discrepancies between simulated and experimental electron heat fluxes and indicates a complex interaction of short and long wavelength turbulence.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

The charge exchange recombination spectroscopy diagnostic on the upgraded Lithium Tokamak eXperiment (LTX-β)

The Lithium Tokamak eXperiment has undergone an upgrade to LTX-β, a major part of which is the addition of neutral beam injection (NBI). NBI has allowed for a new charge exchange recombination spectroscopy (CHERS) system to be installed in order to measure impurity concentrations, ion temperature, and toroidal velocity. Previously on LTX measuring these parameters relied on passive spectroscopy and inversion techniques and had large uncertainty. The CHERS system has 52 total views, split into four groups of 13, half facing toward the beam and half symmetrically facing away from the beam, so the background non-beam related emission can be simultaneously subtracted. Both sets of views sample a major radius of 27-59 cm, with resolution through the beam of 1.5-2.5 cm. LTX-β is expected to have its magnetic axis near 35 cm, with minor radii of 18-23 cm. Three separate spectrometers will be used for the diagnostic, giving the system great flexibility to simultaneously measure emission from multiple impurity lines. The viewing optics are f/1.8, allowing all of the spectrometers to be fully illuminated. Design and calibration of the system as well as the advantages of various configurations of the spectrometers will be highlighted.

Elliott, Drew↗
DOE OSTI2022

System-on-chip integrated circuit technology applications on the DIII-D tokamak for multi-field measurements

Several mm-wave diagnostics on the DIII-D tokamak provide multi-scale and multi-dimensional measurements of plasma profile evolution and turbulence fluctuations. Mm-wave fusion plasma diagnostics that adopt system-on-chip integrated circuit technology can provide better space utilization, flexible installation, and improved sensitivity. In order to further extend this technology for additional fusion facilities with a higher toroidal magnetic field, V-band (55–75 GHz) and F-band (90–140 GHz) chips for Microwave Imaging Reflectometer (MIR) and Electron Cyclotron Emission Imaging (ECEI) instruments are developed and tested in the Davis Millimeter Wave Research Center (DMRC). Current measurement data show that correlation between these SoC-based diagnostic instruments with other state-of-the-art diagnostics enables co-located multi-field turbulence fluctuation measurement.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Avoiding Tokamak Disruptions by Applying Static Magnetic Fields That Align Locked Modes with Stabilizing Wave-Driven Currents [Avoiding Tokamak Disruptions by Magnetically Aligning Locked Modes with Stabilizing Wave-Driven Currents]

The international ITER tokamak has the objective of demonstrating the scientific feasibility of magnetic confinement fusion as a source of energy. A concern towards the achievement of this goal is represented by major disruptions: complete losses of confinement often initiated by a non-rotating ('locked') magnetic island created by magnetic reconnection. During disruptions, energy and particles accumulated in the plasma volume over many seconds are lost in a few milliseconds and released on the plasma-facing materials. In addition, multi-MA level currents flowing in the tokamak plasma for its sustainment and confinement are lost, also in milliseconds, thus terminating the plasma discharge and causing electromagnetic stresses that, if unmitigated, could lead to excessive device wear. Moreover it is shown that magnetic perturbations can be used to avoid disruptions by "guiding" the magnetic island to lock in a position where it is accessible to millimetre wave beams that fully stabilize it.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

DESIGN AND TESTING OF DIMES CARBON ABLATION RODS IN THE DIII-D TOKAMAK

We present the design of ATJ graphite rods developed for ablation experiments under high heat flux (up to 50 MW/m2) in the lower divertor of the DIII-D tokamak, a magnetic plasma confinement device. This work is motivated by the need to test ablation models relevant to carbon-based thermal shields used in high-speed spacecraft atmospheric entries, where the heat fluxes encountered can be comparable to those achieved in the DIII-D divertor plasma. Several different designs for the flowfacing side of the rod are analyzed, including "sharp nose," "blunt," and "concave". The last shape is studied for its potential to lower heat fluxes at the rod surface by increased radiation from trapped neutrals and reduced parallel plasma pressure. We also analyze the possibility of applying a thin (approximately 30 microns) layer of silicon carbide (SiC) to the exposed part of several carbon ablation rods to benchmark its erosion calculations and lifetime predictions. Such calculations are of interest as SiC represents a promising material for both thermal protection systems (TPS) and a fusion plasma-facing material (PFM). Preliminary results from the DIII-D rod ablation experiments are also discussed.

Orlov, Dmitri M.↗
DOE OSTI2015

Reversal of particle flux in collisional-finite beta tokamak discharges

The mixed gradient method [Zhong et al. Phys. Rev. Lett. 111, 265001 (2013)] is adopted and effects of collisions and finite beta are included in the Weiland 9-equation fluid model. The particle flux and particle pinch, obtained using the Weiland anomalous transport fluid model, are compared with Tore Supra experimental results. Particle transport is also studied using predictive simulation data for an experimental advanced superconducting tokamak discharge in which neutral beam heating is utilized. The effects of collisions on particle transport are studied by turning collisions on and off in the Weiland model. It is found that the particle pinch region is related to the mode structure. The particle pinch region coincides with the region where the strong ballooning modes are present due to large gradients. The general properties of the fluid model are examined by finding regions where collisions can enhance the particle pinch.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

First measurements of Hiro currents in vertical displacement event in tokamaks

Specially designed tiles were setup in the 2012 campaign of the Experimental Advanced Superconducting Tokamak (EAST), to directly measure the toroidal surface currents during the disruptions. Hiro currents with direction opposite to the plasma currents have been observed, confirming the sign prediction by the Wall Touching Vertical Mode (WTVM) theory and numerical simulations. During the initial phase of the disruption, when the plasma begins to touch the wall, the surface currents can be excited by WTVM along the plasma facing tile surface, varying with the mode magnitude. The currents are not observed in the cases when the plasma moves away from the tile surface. This discovery addresses the importance of the plasma motion into the wall in vertical disruptions. WTVM, acting as a current generator, forces the Hiro currents to flow through the gaps between tiles. This effect, being overlooked so far in disruption analysis, may damage the edges of the tiles and is important for the ITER device.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

What Can and Cannot Be Expected from Tokamak Fusion

This article is a response to the analysis of nuclear fusion by V. V. Orlov and L. I. Ponomarev in their article “Nuclear problems of thermonuclear power generation” [Atomnaya Energiya, 124, No. 2, 105–110 (2018); Atomic Energy, 124, No. 2, 129–138 (2018)], pointing to the decades-long lack-of-progress in thermonuclear energy research. Here, the authors of the article attribute the lack-of-progress to fundamental problems in the nuclear aspects of thermonuclear fusion. Here, the reason for the current stagnation – essentially, the degeneration of thermonuclear fusion that commenced in the 1950s–1970s with rapid development – is presented. The scientific investigations have been and remain within the scope of the Tamm–Sakharov concept of toroidal plasma confinement, which provided the first impetus to progress but soon showed its insufficiency. An additional element is considered – the suppression of plasma cooling by using moving liquid lithium to pump out plasma. For tokamaks this opens up new prospects for obtaining burning plasma suitable for use in hybrid (fission-fusion) reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Compact formulas for bounce/transit averaging in axisymmetric tokamak geometry

Compact formulas for bounce and transit orbit averaging of the fluctuation-amplitude eikonal factor in axisymmetric tokamak geometry, which is frequently encountered in bounce-gyrokinetic description of microturbulence, are given in terms of the Jacobi elliptic functions and elliptic integrals. These formulas are readily applicable to the calculation of the neoclassical susceptibility in the framework of modern bounce-gyrokinetic theory. In the long-wavelength limit for axisymmetric electrostatic perturbations, we recover the expression for the Rosenbluth-Hinton residual zonal flow [M. N. Rosenbluth and F. L. Hinton, Phys. Rev. Lett. 80, 724 (1998)] accurately.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Effects of the magnetic equilibrium on gyrokinetic simulations of tokamak microinstabilities

The general geometry of the experimental tokamak magnetic equilibrium is implemented in the global gyrokinetic simulation code GEM. Compared to the general geometry, the well used Miller parameterization of the magnetic equilibrium is a good approximation in the core region and up to the top of the pedestal. Linear simulations indicate that results with the two geometries agree for r/a ≤ 0.9. However, in the edge region, the instabilities are sensitive to the magnetic equilibrium in both the L-mode and the H-mode plasmas. A small variation of the plasma shaping parameters leads to large changes to the edge instability.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Numerical investigation of fast-wave propagation and radio-frequency sheath interaction with a shaped tokamak wall

Interactions between propagating fast waves and radio-frequency (RF) sheaths in the ion cyclotron range of frequencies are numerically investigated based on a cold fluid plasma model coupled with a sheath boundary condition. In this two-dimensional study, the capability of the finite element code rfSOL, which was developed in previous numerical work, is extended to analyze self-consistent RF sheath-plasma interaction problems in a tokamak with a non-circular cross-section. It is found that a large sheath voltage is generated near the edges of the limiter-shaped deformation as a result of the conversion from fast to slow waves on the sheaths. The sheath voltage associated with this conversion is particularly significant in the localized region where the contact angle between the magnetic field line and the conducting wall varies rapidly along the curved sheath surface, which is consistent with the results in previous one-dimensional theoretical work. The dependences of the RF sheaths on various parameters in plasma such as the toroidal wavenumber, edge plasma density, and the degree of the RF wave absorption in the core region are also examined in detail.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Braking of tearing mode rotation by ferromagnetic conducting walls in tokamaks

An in-depth investigation of the braking of tearing mode rotation in tokamak plasmas via eddy currents induced in external ferromagnetic conducting structures is performed. In general, there is a “forbidden band” of tearing mode rotation frequencies that separates a branch of high-frequency solutions from a branch of low-frequency solutions. When a high-frequency solution crosses the upper boundary of the forbidden band, there is a bifurcation to a low-frequency solution, and vice versa. The bifurcation thresholds predicted by simple torque-balance theory (which takes into account the electromagnetic braking torque acting on the plasma, as well as the plasma viscous restoring torque, but neglects plasma inertia) are found to be essentially the same as those predicted by more complicated time-dependent mode braking theory (which takes inertia into account). Significant ferromagnetism causes otherwise electromagnetically thin conducting structures to become electromagnetically thick and also markedly decreases the critical tearing mode amplitude above which the mode “locks” to the conducting structures (i.e., the high-frequency to low-frequency bifurcation is triggered). On the other hand, if the ferromagnetism becomes too large, then the forbidden band of mode rotation frequencies is suppressed, and the mode frequency consequently varies smoothly and reversibly with the mode amplitude.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Landau resonant modification of multiple kink mode contributions to 3D tokamak equilibria

Detailed measurements of the plasma's response to applied magnetic perturbations provide experimental evidence that the form of three-dimensional (3D) tokamak equilibria, with toroidal mode number n = 1, is determined by multiple stable kink modes at high-pressure. For pressures greater than the ideal magnetohydrodynamic (MHD) stability limit, as calculated without a stabilizing wall, the 3D structure transitions in a way that is qualitatively predicted by an extended MHD model that includes kinetic wave-particle interactions. These changes in poloidal mode structure are correlated with the proximity of rotation profiles to thermal ion bounce and the precession drift frequencies suggesting that these kinetic resonances are modifying the relative amplitudes of the stable modes. These results imply that each kink may eventually be independently controlled.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for {sup 233}U fuel production: 1. Concept of a stationary Tokamak as a neutron source

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of {sup 233}U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Simple, general, realistic, robust, analytic tokamak equilibria. Part 1. Limiter and divertor tokamaks

Tokamak equilibria have been derived that are analytic solutions to the Grad–Shafranov equation. Here in this paper, Part 1, describes a wide range of such equilibria including smooth limiter surfaces, double- and single-null divertor surfaces, arbitrary aspect ratio, elongation, triangularity and beta. Part 2 generalizes the analysis to include edge pedestals and toroidal flow.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak

New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and T{sub e} monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800–2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma T{sub e}, n{sub e} estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000–1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor T{sub e} monitoring aimed at divertor detachment real-time feedback control.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗