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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 OSTI2024

Using convolutional neural networks to detect edge localized modes in DIII-D from Doppler backscattering measurements

In H-mode tokamak plasmas, the plasma is sometimes ejected beyond the edge transport barrier. These events are known as edge localized modes (ELMs). ELMs cause a loss of energy and damage the vessel walls. Understanding the physics of ELMs, and by extension, how to detect and mitigate them, is an important challenge. In this paper, we focus on two diagnostic methods—deuterium-alpha (D α ) spectroscopy and Doppler backscattering (DBS). The former detects ELMs by measuring Balmer alpha emission, while the latter uses microwave radiation to probe the plasma. DBS has the advantages of having a higher temporal resolution and robustness to damage. These advantages of DBS diagnostic may be beneficial for future operational tokamaks, and thus, data processing techniques for DBS should be developed in preparation. In sight of this, we explore the training of neural networks to detect ELMs from DBS data, using D α data as the ground truth. With shots found in the DIII-D database, the model is trained to classify each time step based on the occurrence of an ELM event. The results are promising. When tested on shots similar to those used for training, the model is capable of consistently achieving a high f1-score of 0.93. Furthermore, this score is a performance metric for imbalanced datasets that ranges between 0 and 1. We evaluate the performance of our neural network on a variety of ELMs in different high confinement regimes (grassy ELM, RMP mitigated, and wide-pedestal), finding broad applicability. Beyond ELMs, our work demonstrates the wider feasibility of applying neural networks to data from DBS diagnostic.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Design Considerations for the Implementation of a High-Field-Side Transient CHI System on QUEST

Transient coaxial helicity injection (T-CHI), a method first developed on the small helicity injected torus (HIT-II) experiment and then validated on the much larger National Spherical Torus Experiment (NSTX) device, is a method to initiate an inductive-like tokamak plasma discharge without reliance on the central solenoid. A CHI discharge is initiated by driving current along magnetic flux that connects the inner and outer divertor plates on one end of the tokamak. To permit this, on both HIT-II and NSTX, toroidal ceramic insulators were used to electrically separate the inner and outer vessel components. The use of such large toroidal vacuum insulators may not be easy to implement in reactors. To address this issue, the Q-shu University experiments with steady-state spherical tokamak (ST) (QUEST) is developing a reactor-relevant CHI configuration in which one of the divertor plates is electrically insulated from the rest of the vessel. The first application of T-CHI on QUEST biased the CHI electrode to the outer vessel. While the CHI discharges could be easily generated, it was found that as the discharge filled the vessel, the separation distance between the injector magnetic flux footprints widened, a condition that is not favorable for the generation of closed flux surfaces. Biasing the electrode to the inner wall is a configuration similar to that used on NSTX and HIT-II, but initial testing in this configuration has proved to be challenging. The design described here overcomes the present limitation by locating the CHI electrode much closer to the CHI injector flux coil and using an NSTX-like gas injection manifold to enable high-field-side T-CHI startup on QUEST. Finally, the concepts described in this article should also benefit the future implementation of T-CHI systems in other tokamaks and spherical tokamaks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Characterization of the ELM-free negative triangularity edge on DIII-D

Tokamak plasmas with strong negative triangularity (NT) shaping typically exhibit fundamentally different edge behavior than conventional L-mode or H-mode plasmas. On DIII-D, every plasma with sufficiently negative triangularity (δ < δ crit ≃ -0.12) is found to be inherently free of edge localized modes (ELMs), even at injected powers well above the predicted L-H power threshold. It is also possible to access an ELM free state at weaker average triangularities provided that at least one of the two x-points is still sufficiently negative. Access to the ELM-free NT scenario is found to coincide with the closure of the second stability region for infinite-n ballooning modes, suggesting that ballooning stability may play a role in limiting the accessible pressure gradient in NT plasmas. Despite this, NT plasmas are able to support small pedestals and are typically characterized by an enhancement of edge pressure gradients beyond those found in traditional L mode plasmas. Further, the pressure gradient inside of this small pedestal is unusually steep, allowing access to high core performance that is competitive with other ELM-free regimes previously achieved on DIII-D. Since ELM-free operation in NT is linked directly to the magnetic geometry, NT fusion pilot plants are predicted to maintain advantageous edge conditions even in burning plasma regimes, potentially eliminating reactor core-integration issues caused by ELMs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

TorbeamNN: machine learning-based steering of ECH mirrors on KSTAR

We have developed TorbeamNN: a machine learning surrogate model for the TORBEAM ray tracing code to predict electron cyclotron heating (ECH) and current drive locations in tokamak plasmas. TorbeamNN provides more than a 100 times speed-up compared to the highly optimized and simplified real-time implementation of TORBEAM without any reduction in accuracy compared to the offline, full fidelity TORBEAM code. The model was trained using KSTAR ECH mirror geometries and works for both O-mode and X-mode absorption. The TorbeamNN predictions have been validated both offline and real-time in experiment. TorbeamNN has been utilized to track an ECH absorption vertical position target in dynamic KSTAR plasmas as well as under varying toroidal mirror angles and with a minimal average tracking error of 0.5 cm.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Disruption Mitigation Solutions for Long Pulse Tokamaks (Final Technical Report)

The project "Disruption Mitigation Solutions for Long Pulse Tokamaks" was a collaborative effort between ORNL, GA, and UCSD. The main goal of the project was to study the injection of shattered cryogenic pellets (SPI) as a method of reducing wall damage from global tokamak plasma instabilities called disruptions. This work was primarily carried out in the JET tokamak in the UK. The team already had extensive experience studying PSI in the medium sized DIII-D tokamak in the US, but adding data from the larger JET tokamak enabled more confident extrapolation to the even larger planned ITER tokamak in France. Each of the collaborating institutions focused on different aspects of SPI disruption mitigation. The UCSD collaboration focused on the volume recombination of post-disruption runaway electron (RE) beams by massive hydrogen injection. Disruptions can sometimes result in the formation of toroidal (ring shaped) beams of highly energetic (relativistic) electron beams, which can cause serious local wall damage, even in present-day tokamaks. A promising method to reduce wall damage from the RE beams is to inject massive amounts of H 2 or D 2 into the beams using SPI. although the precise method by which the injected H 2 reducses RE wall damage is still being investigated, it seems clear that a necessary and required step for the reduction to occur is for the H 2 injection to cause volume recombination (free electron density reduction) of the background cold plasma surrounding the fast electron beam. The UCSD collaboration focused on answering the question of if this recombination could be achievable in the future reactor-size tokamak ITER using realistic levels of H 2 injection. A model capturing the most essential physics was developed and was constrained using both DIII-D and JET data. This model was then used to extrapolate to ITER. It was predicted that RE beam recombination should be achievable in ITER, although in the worst-case scenari (a 10 MA Re beam in ITER), recombination is barely achieved within the maximum allowed levels of H 2 injection.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

The constraint of plasma power balance on runaway avoidance

Abstract In a post-thermal-quench plasma, mitigated or unmitigated, the plasma power balance is mostly between collisional or Ohmic heating and plasma radiative cooling. In a plasma of atomic mixture { n α } with α labeling the atomic species, the power balance sets the plasma temperature, ion charge state distribution { n α i } with i the charge number, and through the electron temperature T e and ion charge state distribution { n α i } , the parallel electric field E ∥ . Since the threshold electric field for runaway avalanche growth E a v is also set by the atomic mixture, ion charge state distribution and its derived quantity, the electron density n e , the plasma power balance between Ohmic heating and radiative cooling imposes a stringent constraint on the plasma regime for avoiding and minimizing runaways when a fusion-grade tokamak plasma is rapidly terminated.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Parametric dependencies of locked mode thresholds in KSTAR L-mode plasmas

Resonant error fields with toroidal harmonics n = 1 and n = 2 in otherwise axisymmetric tokamak plasmas can drive disruptive locked modes when their strength exceeds a threshold value. This paper presents the sensitivity of error field threshold in global parameters such as line-averaged density and toroidal magnetic field as well as toroidal harmonics dependence in KSTAR L-mode plasmas. The error field threshold for mode-locking increases along with density up to a roll-over density, which coincides well with the linear ohmic confinement (LOC) to saturated ohmic confinement (SOC) density for the studied KSTAR plasmas. As a result, this non-monotonic density correlation of the error field threshold in LOC–SOC can be explained by a modification of existing theoretical scaling in SOC, while plasma rotation can also play an important role in setting the resonant error field threshold.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Halo current rotation scaling in post-disruption plasmas

Abstract Halo current (HC) rotation during disruptions can be potentially dangerous if resonant with the structures surrounding a tokamak plasma. We propose a drift-frequency-based scaling law for the rotation frequency of the asymmetric component of the HC as a function of toroidal field strength and plasma minor radius ( f rot ∝ 1/ B T a 2 ). This scaling law is consistent with results reported for many tokamaks and is motivated by the faster HC rotation observed in the HBT-EP tokamak. Projection of the rotation frequency to ITER and SPARC parameters suggest the asymmetric HC rotation will be on the order of 10 Hz and 60 Hz, respectively.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Regulation of the central safety factor and normalized beta under low NBI torque in DIII-D

An algorithm has been designed to simultaneously control the central safety factor (q 0 ) and normalized beta (β N ) while ensuring near-zero torque from the neutral beam injection in DIII-D. Feedback control of q 0 and β N in tokamaks can be beneficial due to the close relationship that these variables have with plasma performance and magneto-hydrodynamic stability. In addition, low neutral-beam-torque conditions are of special interest in present devices because future burning-plasma tokamaks such as ITER will most likely operate at very low plasma rotation. The control synthesis of the algorithm presented in this work is based on a linearized, one-dimensional (1D) model of the current-profile dynamics coupled with a zero-dimensional (0D) plasma-energy balance. The actuators considered are neutral beam injection and electron-cyclotron heating and current drive, and discrete logic determines the neutral-beam injection powers that deliver near-zero torque. Here, the algorithm has been tested in nonlinear, 1D simulations using COTSIM (Control-Oriented Transport SIMulator) and in DIII-D experiments, demonstrating satisfactory performance.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

3D radiated power analysis of JET SPI discharges using the Emis3D forward modeling tool

Abstract Precise values for radiated energy in tokamak disruption experiments are needed to validate disruption mitigation techniques for burning plasma tokamaks like ITER and SPARC. Control room analysis of radiated power ( P rad ) on JET assumes axisymmetry, since fitting 3D radiation structures with limited bolometry coverage is an under-determined problem. In mitigated disruptions, radiation is toroidally asymmetric and 3D, due to fast-growing 3D MHD modes and localized impurity sources. To address this problem, Emis3D adopts a physics motivated forward modeling (‘guess and check’) approach, comparing experimental bolometry data to synthetic data from user-defined radiation structures. Synthetic structures are observed with the Cherab modeling framework and a best fit chosen using a reduced χ 2 statistic. 2D tomographic inversion models are tested, as well as helical flux tubes and 3D MHD simulated structures from JOREK. Two nominally identical pure neon shattered pellet injection (SPI) mitigated discharges in JET are analyzed. 2D tomographic inversions with added toroidal freedom are the best fits in the thermal quench (TQ) and current quench (CQ). In the pre-TQ, 2D reconstructions are statistically the best fits, but are likely over-optimized and do not capture the 3D radiation structure seen in fast camera images. The next-best pre-TQ fits are helical structures that extend towards the high-field side, consistent with an impurity flow under the magnetic nozzle effect also observed in JOREK simulations. Whole-disruption radiated fractions of 0.98 + 0.03 / − 0.29 and 1.01 + 0.02 / − 0.17 are found, suggesting that the stored energy may have been fully mitigated by each SPI, although mitigation efficiencies well below ITER and SPARC requirements for high energy pulses are still within the large uncertainties. Emis3D is also used to validate JOREK SPI simulations, and confirms improvements in matching experiment from changes to impurity modeling. Time-dependent toroidal peaking factors are calculated and discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Impact of Variable Perpendicular Transport Coefficients in WEST Simulations Using SolEdge-HDG

Plasma–wall interaction is one of the key research topics on the way to controlled fusion. To study the best operational designs with reduced heat and particle fluxes onto tokamak plasma facing components (PFCs) comprehensive plasma simulations are required. A recent implementation of a hybridized discontinuous Galerkin scheme into a new version of SolEdge code has the advantage of using magnetic equilibrium-free mesh. This allows us to conduct pioneering 2-D transport simulations of a full discharge in the WEST tokamak. In this work, we implemented plasma transport coefficients as functions of coordinate in the poloidal plane and neutral diffusion as a function of neutral mean free path. Moreover, the perpendicular convection flux terms were added to the code. Using the new features, a few test cases were investigated. Finally, the influence of nonconstant transport coefficients on the simulated particle and heat fluxes onto the WEST tokamak PFCs are demonstrated.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

A simple vacuum suitcase for enabling plasma facing component characterization in fusion devices

We have demonstrated a vacuum suitcase to transport samples in vacuo to a surface analysis station for characterization of tokamak plasma facing components (PFCs). This technique enables surface analysis at powerful, dedicated stations that are not encumbered by design constraints imposed on them by a tokamak. The vacuum suitcase is an alternative solution to characterizing PFCs using diagnostics that are designed and built around a tokamak. The vacuum suitcase, called the Sample Exposure Probe (SEP), features mobile ultra-high vacuum pumping. Active pumping under high vacuum enables sample transfer between the Lithium Tokamak eXperiment-β (LTX-β) and a high resolution X-ray Photoelectron Spectroscopy (XPS) system that is situated close by. A thermocouple inserted in the back of the sample head measures heat flux from the plasma during exposure, and together with a button heater, allows the sample to match the LTX-β PFCs in high temperature operations. As vacuum conditions are better during transfer and analysis than in the tokamak, less contamination is introduced to the samples. XPS scans on a dedicated analysis station enable peak identification due to higher resolution and signal to noise ratio. A similar probe could be implemented for other fusion devices. As a result, the SEP is the first vacuum suitcase implementation for fusion applications that incorporates active pumping.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

Hermes-3: Multi-component plasma simulations with BOUT++

A new open source tool for fluid simulation of multi-component plasmas is presented, based on a flexible software design that is applicable to scientific simulations in a wide range of fields. Hermes-3 is built on plasma simulation framework BOUT++, consolidating earlier SD1D and Hermes models into a single code that can be configured at run-time to solve plasma models in 1D, 2D or 3D, either for transport (steady-state) or turbulent (time-evolving) problems, with an arbitrary number of ion and neutral species. Here, we describe the improved numerical algorithms and software design that have been implemented in Hermes-3. To demonstrate the capabilities of this tool, applications relevant to the boundary of tokamak plasmas are presented: 1D simulations of diveror plasmas evolving equations for all charge states of neon and deuterium; 2D transport simulations of tokamak equilibria in single-null X-point geometry with plasma ion and neutral atom species; and simulations of the time-dependent propagation of plasma filaments (blobs). Hermes-3 is publicly available on Github under the GPL-3 open source license. The repository includes documentation and a suite of unit, integrated and convergence tests.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Investigation of the Suppression of Edge Localized Modes by Resonant Magnetic Perturbations in NSTX/NSTX-U (Final Technical Report)

The purpose of the grant was to investigate the suppression of edge localized modes by resonant magnetic perturbations in both NSTX and NSTX-U. Unfortunately, the project quickly exhausted all of the relevant experimental data from NSTX. It was originally intended to use new data from NSTX-U, when it restarted. However, the considerable delay in the restart made this impossible. Hence, with the permission of DOE, the grant was pivoted toward a more general examination of the MHD stability of tokamak plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Velocity-space distribution function of fast ions in a sawtoothing plasma

This study explores the influence of sawtooth oscillations on the velocity space distribution of fast ions in tokamak plasma discharges. The relevant Fokker–Planck equation for fast ions is solved analytically. Two distinct effects arising from the temperature drop associated with a sawtooth crash and their impact on the distribution function of fast ions are considered. The first effect involves the modulation of the fusion alpha particle source on the timescale of the sawtooth period, linked to the drop in fusion yield resulting from the sawtooth temperature relaxations. The second effect is tied to the increase of the slowing-down time during the sawtooth ramp, causing particles born later in the sawtooth cycle to experience reduced slowing down compared to those born right after the crash, creating an accumulation-like mechanism at higher energies. In regimes where the sawtooth period is shorter than the fast ion slowing-down time, the combined influence of these effects gives rise to fast ion distribution functions that transiently exhibit positive slopes in velocity space.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

The role of shear flow collapse and enhanced turbulence spreading in edge cooling approaching the density limit

Abstract Experimental studies of the dynamics of shear flow and turbulence spreading at the edge of tokamak plasmas are reported. Scans of line-averaged density and plasma current are carried out while approaching the Greenwald density limit on the J-TEXT tokamak. In all scans, when the Greenwald fraction f G = n ¯ / n G = n ¯ / ( I p / π a 2 ) increases, a common feature of enhanced turbulence spreading and edge cooling is found. The result suggests that turbulence spreading is a good indicator of edge cooling, indeed better than turbulent particle transport is. The normalized turbulence spreading power increases significantly when the normalized E × B shearing rate decreases. This indicates that turbulence spreading becomes prominent when the shearing rate is weaker than the turbulence scattering rate. The asymmetry between positive/negative (blobs/holes) spreading events, turbulence spreading power and shear flow are discussed. These results elucidate the important effects of interaction between shear flow and turbulence spreading on plasma edge cooling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Simulations of vertical displacement oscillatory modes and global Alfvén Eigenmodes in JET geometry

Vertical Displacement Oscillatory Modes (VDOM), with frequency in the Alfvén range, are natural modes of oscillation of magnetically confined laboratory plasmas with elongated cross-section. These axisymmetric modes arise from the interaction between the plasma current, which is in equilibrium with currents flowing in external coils, and perturbed currents induced on a nearby conducting wall. The restoring force exerted by these perturbed currents on the vertical motion of the plasma column leads to its oscillatory behavior. An analytic model for VDOM was proposed based on an idealized 'straight tokamak' equilibrium with uniform equilibrium current density. This article introduces the first numerical simulations of VDOM in a realistic JET tokamak configuration, using the extended-MHD code NIMROD and drawing comparisons with Global Alfvén Eigenmodes (GAE). The results show qualitative agreement with analytic predictions regarding mode frequency and radial structure, supporting the identification of VDOM as a fundamental oscillation mode in tokamak plasmas. VDOM and GAE are modeled in a representative JET discharge, where axisymmetric perturbations with toroidal mode number n = 0 driven unstable by fast ions were observed. The two modes are examined separately using a forced oscillator within the NIMROD code, which enables a comparison of their characteristics and helps identify the experimentally observed mode possibly as a GAE.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

A laser inverse Compton scattering diagnostic to study runaway electron dynamics during tokamak disruptions

Both technical and simulation progress has been made toward implementing a laser inverse Compton scattering diagnostic on the DIII-D tokamak. This is a novel diagnostic which uses a short (sub-ns) pulse laser together with a fast (sub-ns) gated x-ray detector to measure relativistic electrons inside a tokamak plasma. The study of these relativistic electrons is important to make sure they can be removed from any tokamak-based power reactor, as they could damage vessel walls and cooling lines. This diagnostic is completely new for magnetic fusion research and has never been implemented before on a tokamak. It is found that good signals and radial profiles could be obtained with an optimized high-energy (7 Joule) laser. Using a standard off-the-shelf Nd:YAG (0.7 Joule) laser would give some signals above the noise but no ability to measure the relativistic electron radial profile.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗