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Results for “CURRENT-DRIVE HEATING”

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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Electron-cyclotron-resonance heating in Wendelstein 7-X: A versatile heating and current-drive method and a tool for in-depth physics studies

For stellarators, which need no or only small amounts of current drive, electron-cyclotron-resonance heating (ECRH) is a promising heating method even for the envisaged application in a fusion power plant. Wendelstein 7-X (W7-X) is equipped with a steady-state capable ECRH system, operating at 140 GHz, which corresponds to the 2nd cyclotron harmonic of the electrons at a magnetic field of 2.5 T. Ten gyrotrons are operational and already delivered 7 MW to W7-X plasmas. Combined with pellet injection, the highest triple product (0.68 × 10 20 keV m –3 s), observed up to now in stellarators, was achieved. For the first time, W7-X plasmas were sustained by 2nd harmonic O-mode heating, approaching the collisionality regime for which W7-X was optimized. Power deposition scans did not show any indication of electron temperature profile resilience. In low-density, low-power plasmas a compensation of the bootstrap current with electron-cyclotron current drive (ECCD) was demonstrated. Furthermore, sufficiently strong ECCD close to the plasma centre produced periodic internal plasma-crash events, which coincide with the appearance of low order rationals of the rotational transform.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Observation of second harmonic electron cyclotron resonance heating and current-drive transition during non-inductive plasma start-up experiment in QUEST

Here, noninductive plasma current start-up using 2nd harmonic electron cyclotron resonance heating (ECRH) with oblique radio frequency (RF) injection is demonstrated in a Q-shu University experiment with steady-state spherical tokamak. A strong transition was observed in the heating and plasma current ramp-up. The initial bulk electron heating regime exhibits T ebulk ~ 140 eV and no hard x-ray (HXR) emission with a low I p of ~15 kA; it abruptly transitions to a regime that exhibits a low T ebulk of ~10 eV and a strong HXR emission with a high I p of ~50 kA. This behavior is distinctly different from that observed in previous fundamental ECRH experiments. The mechanism of the heating and current drive transition are investigated considering wave power absorption and plasma power balance. The results indicate that the transition is caused by the favorable heating of tail electrons where the RF power absorption at the 2nd harmonic increases nearly linearly with T etail , while the power transfer from the tail electrons to the bulk electrons decreases with 1/T etail 0.5 . This causes a rapid transition to a state with high T etail while reducing T ebulk towards colder ion temperature. The understanding of the transition mechanism helps to consider plasma current start-up using 2nd harmonic ECRH for tokamak reactors such as JT-60 SA and ITER.

QUEST↗
DOE OSTI2023

Flexible, integrated modeling of tokamak stability, transport, equilibrium, and pedestal physics

The STEP (Stability, Transport, Equilibrium, and Pedestal) integrated-modeling tool has been developed in OMFIT to predict stable, tokamak equilibria self-consistently with core-transport and pedestal calculations. STEP couples theory-based codes to integrate a variety of physics, including magnetohydrodynamic stability, transport, equilibrium, pedestal formation, and current-drive, heating, and fueling. The input/output of each code is interfaced with a centralized ITER-Integrated Modelling & Analysis Suite data structure, allowing codes to be run in any order and enabling open-loop, feedback, and optimization workflows. This paradigm simplifies the integration of new codes, making STEP highly extensible. STEP has been verified against a published benchmark of six different integrated models. Core-pedestal calculations with STEP have been successfully validated against individual DIII-D H-mode discharges and across more than 500 discharges of the H98,y2 database, with a mean error in confinement time from experiment less than 19%. STEP has also reproduced results in less conventional DIII-D scenarios, including negative-central-shear and negative-triangularity plasmas. Predictive STEP modeling has been used to assess performance in several tokamak reactors. Simulations of a high-field, large-aspect-ratio reactor show significantly lower fusion power than predicted by a zero-dimensional study, demonstrating the limitations of scaling-law extrapolations. STEP predictions have found promising scenarios for an EXhaust and Confinement Integration Tokamak Experiment, including a high-pressure, 80%-bootstrap-fraction plasma. ITER modeling with STEP has shown that pellet fueling enhances fusion gain in both the baseline and advanced-inductive scenarios. Finally, STEP predictions for the SPARC baseline scenario are in good agreement with published results from the physics basis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Non-resonant destabilization of (1/1) internal kink mode by suprathermal electron pressure

New experimental observations are reported on the structure and dynamics of short-lived periodic (1, 1) “fishbone”-like oscillations that appear during radio frequency heating and current-drive experiments in tokamak plasmas. For the first time, measurements can directly relate changes in the high energy electrons to the mode onset, saturation, and damping. In the relatively high collisionality of Alcator C-Mod with lower hybrid current drive, the instability appears to be destabilized by the non-resonant suprathermal electron pressure—rather than by wave-particle resonance, rotates toroidally with the plasma and grows independently of the (1, 1) sawtooth crash driven by the thermal plasma pressure.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Modification of the collective Thomson scattering radiometer in the search for parametric decay on TEXTOR

Strong scattering of high-power millimeter waves at 140 GHz has been shown to take place in heating and current-drive experiments at TEXTOR when a tearing mode is present in the plasma. The scattering signal is at present supposed to be generated by the parametric decay instability. Here we describe the heterodyne detection system used to characterize the newly discovered signal measured at TEXTOR, and we present spectral shapes in which the signal can appear under different conditions. The radiation is collected by the receiver through a quasi-optical transmission line that is independent of the electron cyclotron resonance heating transmission line, and so the scattering geometry is variable. The signal is detected with 42 frequency channels ranging from 136 to 142 GHz. We demonstrate that the large signal does not originate from gyrotron spurious radiation. The measured signal agrees well with independent backscattering radiometer data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Diagnostics for machine protection of DEMO

DEMO aims to (i) integrate, demonstrate and validate all relevant technology necessary to convert fusion energy to electrical energy and (ii) that the machine and its operations are economically and environmentally acceptable. To maintain the efficiency and availability of the machine, there are several physics and combined physics/technology issues as well as the engineering issues. Machine Protection (also known as Protection of Investment) addresses both the risks to plant (to avoid costly repair or replacement) and the risks to normal operating time (to avoid loss of productivity and the return on investment). The plasma-related Machine Protection issues involve measurement and control of plasma stability, plasma purity, and plasma-wall interactions. Machine Protection aims to avoid hitting catastrophic limits by using early warning alarm systems, and controlled termination or avoidance, involving coordinated actions of the magnets, gas and auxiliary heating or current-drive systems. This article outlines the key processes, some of which are used in present-day tokamaks and some of which are new specifically for DEMO (e.g. First wall and divertor power handling) and reveals the need to research and develop new science and technology for Machine Protections in DEMO's high radiation and thermal fields. This work was funded by the RCUK Energy Programme under grant EP/I501045 and the European Communities under the contract of Association between EURATOM and CCFE and conducted partly under EFDA PPPT (WP13-DAS04). The views and opinions expressed herein do not necessarily reflect those of the European Commission.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Robust control of the current profile and plasma energy in EAST

Integrated control of the toroidal current density profile, or alternatively the q-profile, and plasma stored energy is essential to achieve advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamics stability, and noninductively driven plasma current. The q-profile evolution is closely related to the evolution of the poloidal magnetic flux profile, whose dynamics is modeled by a nonlinear partial differential equation (PDE) referred to as the magnetic-flux diffusion equation (MDE). The MDE prediction depends heavily on the chosen models for the electron temperature, plasma resistivity, and non-inductive current drives. To aid control synthesis, control-oriented models for these plasma quantities are necessary to make the problem tractable. However, a relatively large deviation between the predictions by these control-oriented models and experimental data is not uncommon. For this reason, the electron temperature, plasma resistivity, and non-inductive current drives are modeled for control synthesis in this work as the product of an “uncertain” reference profile and a nonlinear function of the different auxiliary heating and current-drive (H&CD) source powers and the total plasma current. The uncertainties are quantified in such a way that the family of models arising from the modeling process is able to capture the q-profile and plasma stored energy dynamics from a typical EAST shot. A control-oriented nonlinear PDE model is developed by combining the MDE with the “uncertain” models for the electron temperature, plasma resistivity, and non-inductive current drives. This model is then rewritten into a control framework to design a controller that is robust against the modeled uncertainties. The resulting controller utilizes EAST's H&CD powers and total plasma current to regulate the q profile and plasma stored energy even when mismatches between modeled and actual dynamics are present. The effectiveness of the controller is demonstrated through nonlinear simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Parasitic signals in the receiving band of the Sub-Harmonic Arc Detection system on JET ICRF Antennas

When testing the SHAD system on JET ICRF antennas, parasitic signals in the detection band (5-20MHz) were detected. We have identified emission from grid breakdown events in the Neutral Beam injectors, and Ion Cyclotron Emission from the plasma. Spurious signals in the band 4-10 MHz are also often observed at the onset of ELM events. Such parasitic signals could complicate the design and operation of SHAD in ICRF systems for fusion devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Stray RF Power Estimates From EC Exploitation During ITER Plasma Operations

The EC H and CD system of ITER tokamak is an essential tool for all the phases of ITER operation. Different levels of EC power are required through all the plasma discharge: up to 6.7 MW for assisting the breakdown and burn through, up to 20 MW for current drive and saw-teeth control from the equatorial launcher and up to 20 MW for NTM stabilization from the upper launchers. The assistance to breakdown and burn through is characterized by a very low (if not negligible) RF power absorption by the plasma. A significant level of stray radiation may also arise from partial absorption due to non-optimal plasma parameters and/or wrong injected polarization. The stray power radiated in the vacuum chamber is estimated as a first step toward mitigating potential harmful consequence to in-vessel structures and diagnostics. Power loading of the chamber walls (peek power and average power density for straight beam propagation in the empty chamber) and diffuse stray radiation effects are simulated to infer suitable strategies to avoid damage to first wall and to microwave sensitive components.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Modeling of Synergy Between 4th and 6th Harmonic Absorptions of Fast Waves on Injected Beams in DIII-D Tokamak

In recent moderate to high harmonic fast wave heating and current drive experiments in DIII-D, a synergy effect was observed when the 6{sup th} harmonic 90 MHz fast wave power is applied to the plasma preheated by neutral beams and the 4{sup th} harmonic 60 MHz fast wave. In this paper, we investigate how the synergy can occur using ORBIT-RF coupled with AORSA. Preliminary simulations suggest that damping of 4{sup th} harmonic FW on beam ions accelerates them above the injection energy, which may allow significant damping of 6{sup th} harmonic FW on beam ion tails to produce synergy.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Studying fast wave propagation and absorption at any cyclotron harmonic using a 2D finite element area coordinates wave equation solver

Fourier analysis in the poloidal direction is a standard ingredient in present-day 2D wave equation solvers describing radio frequency waves in hot tokamak plasmas. Although a powerful and elegant technique, Fourier analysis has the disadvantage that a large number of modes is needed to describe the field pattern on a magnetic surface if a short wavelength mode exists on any - even very small - subpart of the particle trajectory. The present paper examines the potential of a method that does not suffer from this drawback: a finite element technique relying on simple linear or cubic area base functions that are defined on irregular elementary surfaces of triangular shape. The wave equation is solved in its weak Galerkin variational form and for realistic 2D tokamak geometry, accounting for the toroidal curvature but assuming the toroidal angle is ignorable, allowing to study the wave pattern for each of the independent toroidal modes excited by the antenna individually.The locally uniform full hot plasma dielectric tensor to all orders in finite Larmor radius was adopted. As the main intended application is the study of fast wave behavior (heating and current drive) at arbitrary harmonics, the wave vector complex amplitude appearing in the dielectric tensor is determined through a local dispersion root evaluation. High frequency fast wave propagation and damping is provided as an illustration in view of possible application of this type of current drive in future high density reactor-like tokamaks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Lower hybrid current drive in a high density diverted tokamak

Experimental observations of LHCD at high density (n-bar{sub e}>10{sup 20}m{approx}{sup 3}) on the Alcator C-Mod tokamak are presented in this paper. Bremsstrahlung emission from relativistic fast electrons in the core plasma drops sharply in single null discharges well below the density limit previously observed on limited tokamaks ({omega}/{omega}{sub LH}{approx}2). Modeling and experimental evidence suggest that the absence of LH driven fast electrons at high density may be due to collisional absorption in the scrape off layer. Experiments show that the expected current drive density dependence is recovered for inner wall limited discharges across the range of densities scanned (0.5x10{sup 20} m{sup -3}<n-bar{sub e}<1.5x10{sup 20} m{sup -3}). Increasing T{sub e} in the periphery of the plasma (0.8<r/a<1.0) also results in a modest increase in non-thermal electron emission at high n-bar{sub e}. Ray tracing/Fokker-Planck simulations of these discharges predict the observed sensitivity to plasma position when the effects of collisional absorption in the SOL are included in the model.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Bremsstrahlung emission modelling and application to fast electron physics

Measurements of fast electron bremsstrahlung (FEB) using hard X-ray (HXR) cameras provide a unique diagnostic for the fast electron dynamics, with resolution in space, time and energy. A synthetic HXR diagnostic is developed to calculate the distribution function and FEB emission, such that experimental results can be directly compared with simulations. Using this comparison, the role of fast electron radial transport in radio-frequency current drive is investigated. For two cases studied in this paper, the HXR emission is consistent with negligible radial transport, although in one case this is in disagreement with the measured driven current.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Electron Cyclotron / Bernstein Wave Heating and Current Drive Experiments using Phased-array Antenna in QUEST

The phased-array antenna system for Electron Cyclotron/Bernstein Wave Heating and Current Drive experiments has been developed in the QUEST. The antenna was designed to excite a pure O-mode wave in the oblique injection for the O-X-B mode conversion experiments, and its good performances were confirmed at a low power level. The plasma current (<{approx}15 kA) with an aspect ratio of 1.5 was started up and sustained by only RF injection in the low-density operations. The long pulse discharge of 10 kA was also attained for 37 s. The new density window to sustain the plasma current was observed in the high-density plasmas. The single-null divertor configuration with the high plasma current (<{approx}25 kA) was attained in the 17 s plasma sustainment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Production of High Energy Tail Electrons by Electron Bernstein Waves during the Current Start-up Discharges in the LATE Device

Toroidal plasma current is started and ramped up by injecting microwave power in the electron cyclotron range of frequency without induction in the LATE device. Radial scanning with hard X-ray pulse height analysis reveals the production of high energy electrons with average energy {approx}100 keV in the radial region from R = 28 cm to 40.5 cm, which are heated by electron Bernstein wave. The radial profile of photon counts in the energy range from 25 to 200 keV is very similar to that of perpendicular pressure obtained by magnetic measurement and equilibrium analysis, suggesting that a significant portion of trapped electrons exists outside the last closed flux surface. The plasma current inside the LCFS is carried mainly by passing electrons, while some portion of the outside current may be generated as a result of the toroidal precession of trapped electrons.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Towards identifying the mechanisms underlying field-aligned edge-loss of HHFW power on NSTX

Fast-wave heating will be a major heating scheme on ITER, as it can heat ions directly and is relatively unaffected by the large machine size unlike neutral beams. However, fast-wave interactions with the plasma edge can lead to deleterious effects such as, in the case of the high-harmonic fast-wave (HHFW) system on NSTX, large losses of fast-wave power in the scrape off layer (SOL) under certain conditions. In such scenarios, a large fraction of the lost HHFW power is deposited on the upper and lower divertors in bright spiral shapes. The responsible mechanism(s) has not yet been identified but may include fast-wave propagation in the scrape off layer, parametric decay instability, and RF currents driven by the antenna reactive fields. Understanding and mitigating these losses is important not only for improving the heating and current-drive on NSTX-Upgrade but also for understanding fast-wave propagation across the SOL in any fast-wave system. This talk summarizes experimental results demonstrating that the flow of lost HHFW power to the divertor regions largely follows the open SOL magnetic field lines. This lost power flux is relatively large close to both the antenna and the last closed flux surface with a reduced level in between, so the loss mechanism cannot be localized to the antenna. At the same time, significant losses also occur along field lines connected to the inboard edge of the bottom antenna plate. The power lost within the spirals is roughly estimated, showing that these field-aligned losses to the divertor are significant but may not account for the total HHFW loss. To elucidate the role of the onset layer for perpendicular fast-wave propagation with regards to fast-wave propagation in the SOL, a cylindrical cold-plasma model is being developed. This model, in addition to advanced RF codes such as TORIC and AORSA, is aimed at identifying the underlying mechanism(s) behind these SOL losses, to minimize their effects in NSTX-U, and to predict their importance in ITER.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2025

WEST L-mode record long pulses guided by predictions using Integrated Modeling

A new record was set on the WEST Tokamak, designed to operate long duration plasmas in a tungsten (W) environment, with an injected energy of 1.15 GJ and a plasma duration 364s. Scenario development was supported by integrated modeling using the High Fidelity Plasma Simulator (HFPS), the European IMAS-coupled version of JETTO/JINTRAC, which integrates physics-driven modules into a unified framework. In particular, a reduced model for Lower-Hybrid heating and Current-Drive (LHCD) and the quasi-linear turbulent transport model TGLF are crucial for long pulses predictions up to the Last Closed Flux Surface (LCFS). Using this workflow, a 100 s reference discharge was modeled and plasma kinetic profiles and loop voltage were quantitatively well matched. In preparation for the recent long duration experiments, non-inductive current-drive actuators (I P , n e , P LHCD ) were varied to determine the operational domain going towards fully non-inductive discharges. In particular, decreasing the plasma current is shown to ease the access to such conditions, with a careful monitoring of (n e , P LHCD ) to avoid machine limitations. In addition, post-prediction experiments conducted within the investigated parameter range validated the predicted dependencies and were shown to be in quantitative agreement. Exploratory work on the use of ECCD for MHD stability purpose is also introduced.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

FY-2015 FES (Fusion and Energy Sciences) Joint Research Target: Final Report for the Period October 1st, 2014, through September 30th, 2015

Annual JRT-15 Target: Conduct experiments and analysis to quantify the impact of broadened current and pressure profiles on tokamak plasma confinement and stability. Broadened pressure profiles generally improve global stability but can also affect transport and confinement, while broadened current profiles can have both beneficial and adverse impacts on confinement and stability. This research will examine a variety of heating and current drive techniques in order to validate theoretical models of both the actuator performance and the transport and global stability response to varied heating and current drive deposition.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗