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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels of isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.

Holtrop, Kurt↗
DOE OSTI2016

Preliminary consideration of CFETR ITER-like case diagnostic system

Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Four-dimensional calibration turntable of the motional Stark effect diagnostic on EAST

The motional Stark effect (MSE) diagnostic is applied to measure the safety factor q and current density profile of a tokamak device, which are important parameters in realizing the high-performance and long-pulse steady state of a tokamak. A single-channel MSE diagnostic based on dual photoelastic modulators, whose sightline meets with the neutral beam injection at a major radius of R = 2.12 m, has been built for the D window of the Experimental Advanced Superconducting Tokamak (EAST). According to the requirements of MSE diagnostic polarimetric calibration, a high-precision four-dimensional calibration turntable, driven by four stepping motors and controlled by software running on the computer, was designed for EAST. The turntable allows us to rapidly calibrate the MSE diagnostic in a series of positions and angles during EAST maintenance. The turntable can move in four dimensions of translation, yaw, pitch, and roll of the polarizer and can create linearly polarized light at any given angle with accuracy of ~0.05° for the MSE system offline calibration. As a result, the experimental results of the MSE diagnostic calibration in the laboratory show that the turntable has the advantages of high positioning accuracy, flexible spatial movement, and convenient control and fully meets the calibration requirements of an MSE diagnosis system.

47 OTHER INSTRUMENTATION↗
DOE OSTI2017

Upgraded flowing liquid lithium limiter for improving Li coverage uniformity and erosion resistance in EAST device

In this paper, we report on design and technology improvements for a flowing liquid lithium (FLiLi) limiter inserted into auxiliary heated discharges in the experimental advanced superconducting tokamak device. In order to enhance Li coverage uniformity and erosion resistance, a new liquid Li distributor with homogenous channels was implemented. In addition, two independent electromagnetic pumps and a new horizontal capillary structure contributed to an improvement in the observed Li flow uniformity (from 30% in the previous FLiLi design to >80% in this FLiLi design). To improve limiter surface erosion resistance, hot isostatic press technology was applied, which improved the thermal contact between thin stainless steel protective layers covering the Cu heat sink. The thickness of the stainless steel layer was increased from 0.1 mm to 0.5 mm, which also helped macroscopic erosion resilience. Finally, despite the high auxiliary heating power up to 4.5 MW, no Li bursts were recorded from FLiLi, underscoring the improved performance of this new design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Simulations of divertor plasmas with inverse sheaths

The effect of strong electron emission from material surfaces has been proposed to form an “inverse sheath”: a region with a positive potential relative to the near-wall plasma which prevents the flow of ions to the wall [M. D. Campanell, “Negative plasma potential relative to electronemitting surfaces,” Phys. Rev. E. 88, 033103 (2013); M. D. Campanell and M. V. Umansky, “Strongly emitting surfaces unable to float below plasma potential,” Phys. Rev. Lett. 116, 1–5 (2016); M. D. Campanell and G. R. Johnson, “Thermionic cooling of the target plasma to a sub-ev temperature,” Phys. Rev. Lett. 122, 1–5 (2019)]. We assess the viability of this regime in a tokamak device using the 2D edge plasma transport code UEDGE [T. Rognlien et al ., “A fully implicit, time dependent 2-D fluid code for modeling tokamak edge plasmas,” J. Nucl. Mater. 196–198, 347–351 (1992)]. Since the UEDGE code does not consider the sheath region directly, we apply boundary conditions at the divertor targets which emulate the physics of both “standard” and “inverse” sheath regimes [R. Masline et al ., “Influence of the inverse sheath on divertor plasma performance in tokamak edge plasma simulations,” Contrib. Plasma Phys. 60, e201900097 (2020)]. Using these boundary conditions, we perform scoping studies to assess plasma parameters near the target by varying the density at the core-edge interface. We observe a smooth transition in the resultant profiles of plasma parameters for the standard sheath, and a bifurcation across the simulation set for plasmas with an inverse sheath. The cause of this bifurcation is assessed by performing the parameter scan both with and without impurity radiation; we observe that the bifurcation persists in both cases, indicating that this bifurcation is caused by plasma recombination.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Inferring the scrape-off layer heat flux width in a divertor with a low degree of axisymmetry

Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MWm -2 . The magnitude of these heat loads is set by the width of the channel, the ‘scrape-off layer’ (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field (≥1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field (≥0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic geometry of the plasma to infer the width of the SOL, with both Eich and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, along with increased heat loads observed locally around the edges of the tiles. Future work in characterising the impact of tile misalignment and uncertainties in the reconstructed divertor magnetic geometry is required in order to further understand the observed heat flux patterns, as are additional investigations into the role potentially being played by an inhomogeneous sheath electric field.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Potential design problems for ITER fusion device

The international thermonuclear experimental reactor (ITER) is a worldwide project currently being built in France for the demonstration of the feasibility of thermonuclear technologies for future realization of successful commercial fusion energy. ITER is of the tokamak based design using strong magnetic fields to confine the very hot plasma needed to induce the fusion reaction. Tokamak devices are currently the front leading designs. Building a successful magnetic fusion device for energy production is of great challenge. A key obstacle to such design is the performance during abnormal events including plasma disruptions and so-called edge-localized modes (ELMs). In these events, a massive and sudden release of energy occurs quickly, due to loss of full or partial plasma confinement, leading to very high transient power loads on the reactor surface boundaries. A successful reactor design should tolerate several of these transient events without serious damages such as melting and vaporization of the structure. This paper highlights, through comprehensive state-of-the-art computer simulation of the entire ITER interior design during such transient events, e.g., ELMs occurring at normal operation and disruptions during abnormal operation, potential serious problems with current plasma facing components (PFCs) design. The HEIGHTS computer package is used in these simulations. The ITER reactor design was simulated in full and exact 3D geometry including all known relevant physical processes involved during these transient events. The current ITER divertor design may not work properly and may requires significant modifications or new innovative design to prevent serious damage and to ensure successful operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Drift kinetic effects on the plasma response in high beta spherical tokamak experiments

High $\beta$ plasma response to the rotating n=1 external magnetic perturbations is numerically studied and compared with National Spherical Torus eXperiment (NSTX). The hybrid magnetohydrodynamic(MHD)-kinetic modeling shows the drift kinetic effects are important to resolve the disagreement of plasma response between the ideal MHD prediction and the NSTX experimental observation when plasma pressure reaches and exceeds the no-wall limit [F. Troyon et al., Plasma Phys. Control. Fusion \textbf{26}, 209 (1984)]. Since the external rotating fields and high plasma rotation are presented in NSTX experiments, the importance of resistive wall effect and plasma rotation on determining the plasma response is also identified, where the resistive wall suppresses the plasma response through the wall eddy current. The inertial energy, due to plasma rotation, destabilizes the plasma. The complexity of plasma response, in this study, indicates that MHD modeling, including comprehensive physics e.g. the drift kinetic effects, resistive wall and plasma rotation, is essential to reliably predict the plasma behavior in high beta spherical tokamak device.

Tokamak↗
DOE OSTI2021

Edge and scrape-off layer modeling for a Fusion Nuclear Science Facility with tungsten walls; a summary report for 2019-21

This report summarizes model development and simulations for the edge/scrape-off layer (SOL) region of a Fusion Nuclear Science Facility (FNSF) as part of the DOE Fusion Energy Systems Studies project. An overview of the FNSF device is given in Ref. 1. Our earlier related modeling of FNSF in the 2015-16 timeframe is reported in Ref. 2, and similar work on the ARIES ACT-1 tokamak device is described in Ref. 3. During 2017-18, we contributed to the analysis of a liquid lithium wall for FNSF [4].

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Helical core reconstruction of a DIII-D hybrid scenario tokamak discharge

Our paper presents the first fully 3-dimensional (3D) equilibrium reconstruction of a helical core in a tokamak device. Using a new parallel implementation of the Variational Moments Equilibrium Code (PARVMEC) coupled to V3FIT, 3D reconstructions can be performed at resolutions necessary to produce helical states in nominally axisymmetric tokamak equilibria. In a flux pumping experiment performed on DIII-D, an external n=1 field was applied while a 3/2 neoclassical tearing mode was suppressed using ECCD. The externally applied field was rotated past a set of fixed diagnostics at a 20 Hz frequency. Furthermore, the modulation, were found to be strongest in the core SXR and MSE channels, indicates a localized rotating 3D structure locked in phase with the applied field. Signals from multiple time slices are converted to a virtual rotation of modeled diagnostics adding 3D signal information. In starting from an axisymmetric equilibrium reconstruction solution, the reconstructed broader current profile flattens the q-profile, resulting in an m=1, n=1 perturbation of the magnetic axis that is $\sim 50\times $ larger than the applied n=1 deformation of the edge. Error propagation confirms that the displacement of the axis is much larger than the uncertainty in the axis position validating the helical equilibrium.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Tokamak divertor plasma emulation with machine learning

Abstract Future tokamak devices that aim to create conditions relevant to power plant operations must consider strategies for mitigating damage to plasma facing components in the divertor. One of the goals of MAST-U tokamak operations is to inform these considerations by researching advanced divertor configurations that aid stable plasma detachment. Machine design, scenario planning and detachment control would all greatly benefit from tools that enable rapid calculation of scenario-relevant quantities given some input parameters. This paper presents a method for generating large, simulated scrape-off layer data sets, which was applied to generate a data set of steady-state Hermes-3 simulations of the MAST-U tokamak. A machine learning model was constructed using a Bayesian approach to hyperparameter optimisation to predict diagnosable output quantities given control-relevant input features. The resulting best-performing model, which is based on a feedforward neural network, achieves high accuracy when predicting electron temperature at the divertor target and carbon impurity radiation front position and runs in around 1 ms in inference mode. Techniques for interpreting the predictions made by the model were applied, and a high-resolution parameter scan of upstream conditions was performed to demonstrate the utility of rapidly generating accurate predictions using the emulator. This work represents a step forward in the design of machine learning-driven emulators of tokamak exhaust simulation codes in operational modes relevant to divertor detachment control and plasma scenario design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

DOE FES FY2017 Joint Research Target Fourth Quarter Milestone Report for theNational Spherical Torus Experiment Upgrade.

A successful high-performance plasma operation with a radiative divertor has been demonstrated on many tokamak devices, however, significant uncertainty remains in accurately modeling detachment thresholds, and in how detachment depends on divertor geometry. Whereas it was originally planned to perform dedicated divertor experiments on the National Spherical Tokamak Upgrade to address critical detachment and divertor geometry questions for this milestone, the experiments were deferred due to technical difficulties. Instead, existing NSTX divertor data was summarized and re-analyzed where applicable, and additional simulations were performed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Runaway electrons in tokamaks

The generation of runaway electrons is a complex and important phenomenon that impacts many areas of plasma physics. Due to the decrease of electron collision frequency with increasing velocity, electrons under strong electric field can experience unlimited “runaway” acceleration. In tokamaks, runaway electrons can be produced in disruptions, due to the strong inductive electric field formed as the thermal energy of plasma gets rapidly lost. This population of runaway electrons can undergo an exponential growth, denoted the runaway electron avalanche, due to hard collisions between relativistic runaway electrons and low energy electrons. It is predicted that in a large tokamak device like the International Thermonuclear Experimental Reactor (ITER), a runway electron beam generated in a disruption event can potentially cause severe damage to the device, which poses a significant challenge for ITER to achieve its mission. It is therefore extremely important to seek an effective mitigation mechanism for runaway electrons. Experimental efforts have been made to study the properties of runaway electrons in tokamaks, including their generation, diffusion, and radiation. In order to understand these experimental results, extensive theoretical and simulation studies of runaway electron physics are required. The main topic of this thesis is to study the wave particle interaction associated with runaway electron beams in tokamaks. The runaway electrons can emit and absorb electromagnetic waves through resonances, and can be diffused in momentum space by the waves. Initially, we address the Cherenkov radiation of runaway electrons, which originates from the polarization of the plasma medium. The energy and momentum loss of the Cherenkov radiation can be modeled by adding a correction to the Coulomb logarithm in the collisional drag force. Subsequently, we address pitch angle scattering caused by normal modes in the plasma, which are driven unstable by the anisotropicity of the runaway electron beam. The fluctuating electromagnetic fields are found to act as a seed for the unstable normal modes. Numerical simulations show that the pitch angle scattering effect from the normal modes, mainly whistler waves, can be significantly larger than that from collisional pitch angle scattering. Finally, we present a synthetic diagnostic tool we developed to calculate the electron cyclotron emission (ECE) from the runaway electrons, and successfully reproduce the prompt growth of the ECE signal observed in DIII-D quiescent runaway electron (QRE) experiments. Within the thesis, we also present the application of the adjoint method to runaway electron research, and show the calculations of the runaway probability function (RPF) and the expected loss time (ELT). These calculations not only help depict the dynamics of runaway electrons in momentum space, but also can be used to efficiently calculate experimentally relevant quantities such as the critical electric field for runaway electron avalanche and the avalanche growth rate.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Scattering of lower hybrid radio frequency waves by cylindrical turbulent structures in the plasma edge in tokamaks

Lower hybrid (LH) radio frequency (RF) waves are used in fusion tokamak devices to generate non-inductively toroidal currents. LH waves are effective in imparting toroidal momentum to electrons in the core of the confined plasma. Since the LH waves are generated by wave guide structures near the wall of a tokamak, the waves have to propagate through a turbulent plasma in the edge region before coupling to the core. It is especially important to quantify the effect of this plasma region on the propagation characteristics of the LH waves, since fusion reactors like ITER will have an extended edge region. Structures in the plasma edge like filaments and blobs which have highly varying density fluctuations compared to the background density, span radial spatial scales that are comparable to the LH wavelength. We study the scattering of the LH waves by the filamentary structures using Maxwell's equations, in which the plasma permittivity is given by the cold plasma dispersion tensor. The collisional absorption of LH waves in the edge region is included by a modification to the elements of the dispersion tensor. The filaments are assumed to be cylindrical with the axis predominantly aligned along the direction of the toroidal magnetic field. Our studies are both analytical and numerical [1,2] and show that these structures can lead to reflection, refraction, diffraction, and side-scattering of both an incident LH plane wave and a Gaussian beam. We will present a variety of different density variations and collisional absorption rates in plasmas with filamentary structures of varying sizes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Stability of drift-cyclotron loss-cone waves in H-mode plasmas

The drift-cyclotron loss-cone mode was first studied in mirror machines. In such devices, particles with small pitch angles are not confined, creating a hole in the velocity distribution function that is a source of free energy and leads to micro-instabilities in the cyclotron-range of frequencies. In the edge region of tokamak devices operating under H-mode conditions, ion loss also occurs. In this case, gradient drift carries ions moving opposite to the plasma current preferentially into the divertor, creating a one-sided loss cone. A simple analysis shows that for the quiescent H-mode plasmas in DIII-D the critical gradient for instability is exceeded within 2 cm of the separatrix, and the maximum growth rate at the separatrix is 3×10 7 s -1 .

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

A review of lithium application for the plasma-facing material in EAST Tokamak

As a low-Z material, lithium is widely tested and used in tokamak devices for particle recycling and impurity control to improve plasma performance. Lithium application as plasma-facing materials via different methods, such as coating, injection and flowing liquid limiter, has been investigated more than 10 years on EAST. This study presents the development of the lithium techniques applied on EAST, and reviews the progresses on the investigation of its interaction with plasma and its effects on the improvement on the plasma performances. In addition, this paper further discusses the potential applications of lithium technique in future fusion devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Fast ion energy distribution from third harmonic radio frequency heating measured with a single crystal diamond detector at the Joint European Torus

Neutron spectroscopy measurements with a single crystal diamond detector have been carried out at JET, for the first time in an experiment aimed at accelerating deuterons to MeV energies with radio frequency heating at the third harmonic. Data are interpreted by means of the expected response function of the detector and are used to extract parameters of the highly non-Maxwellian distribution function generated in this scenario. A comparison with observations using a time of flight and liquid scintillator neutron spectrometers is also presented. The results demonstrate the capability of diamond detectors to contribute to fast ion physics studies at JET and are of more general relevance in view of the application of such detectors for spectroscopy measurements in the neutron camera of next step tokamak devices.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Magnetic-Confinement Fusion—Plasma Theory: Tokamak Magnetohydrodynamic Equilibrium and Stability

Magnetohydrodynamics (MHD) provides a useful model to describe the crucial plasma macroscopic equilibrium and stability behaviors in toroidal tokamak devices by considering the plasma as a conducting fluid interacting with a surrounding confining electromagnetic field. MHD is the most basic plasma model, incorporating most large-scale phenomena, including plasma equilibrium and all major instabilities. MHD equations are obtained by taking moments of the Boltzmann equations for different plasma species. They provide a set of comprehensive physics constrains to compute and optimize the equilibrium plasma shape and pressure and current profiles that are critical to its stability and performance. In the ideal case, the equations have special properties that lead to efficient numerical calculation schemes, the most important of which is the ideal MHD energy principle for linear stability against small departures from equilibrium. In a tokamak plasma, equilibrium pressure is mostly destabilizing for MHD modes, whereas equilibrium current is also often a major driving force. Plasma resistivity creates new freedom for a MHD instability to grow, but there are also cases where the plasma resistivity plays a stabilizing role. Equilibrium toroidal flow and/or flow shear can affect MHD instabilities. Principal MHD instabilities include the internal kink mode, sawtooth, fishbone, external kink, resistive wall mode, resistive interchange, tearing and neoclassical tearing modes (NTMs), locked modes, toroidal Alfven eigenmodes (TAEs), and edge localized modes (ELMs). Fast-growing MHD instabilities can lead to an abrupt plasma disruption and termination that can potentially damage the device plasma facing components (PFCs) and in-vessel structures. Furthermore, an important MHD application is to develop robust techniques to mitigate and control MHD instabilities.

Dispersive pellet injection↗