A5082442710- Magnetic confinement fusion research
- Superconducting Materials and Applications
- Fusion materials and technologies
- Particle accelerators and beam dynamics
- Ionosphere and magnetosphere dynamics
- Plasma Diagnostics and Applications
- Laser-Plasma Interactions and Diagnostics
- Nuclear reactor physics and engineering
- Advanced Data Storage Technologies
- Semiconductor materials and devices
- Advancements in Semiconductor Devices and Circuit Design
- Electrostatic Discharge in Electronics
- Power System Optimization and Stability
- Iterative Learning Control Systems
- Target Tracking and Data Fusion in Sensor Networks
- Radiation Effects in Electronics
- Computational Fluid Dynamics and Aerodynamics
- Particle Detector Development and Performance
- Nuclear Physics and Applications
- Frequency Control in Power Systems
- Fluid Dynamics and Turbulent Flows
- Laser-induced spectroscopy and plasma
- Fault Detection and Control Systems
- Silicon Carbide Semiconductor Technologies
- Wind and Air Flow Studies
General Atomics (United States)
2015-2024
DIII-D National Fusion Facility
2005-2023
Leonardo (United Kingdom)
2018-2023
University of Dundee
2023
Fusion (United States)
2004-2023
Predictive Science (United States)
2019
Newcastle University
2019
University of Surrey
2019
Center for Effective Philanthropy
2017
Max Planck Institute for Plasma Physics
2016
Progress in the area of MHD stability and disruptions, since publication 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137–2664), is reviewed. Recent theoretical experimental research has made important advances both understanding control tokamak plasmas. Sawteeth are anticipated baseline ELMy H-mode scenario, but tools exist to avoid or them through localized current drive fast ion generation. Active other instabilities will most likely be also required ITER. Extrapolation from...
An overview of the present status research toward final design ITER disruption mitigation system (DMS) is given. The DMS based on massive injection impurities, in order to radiate plasma stored energy and mitigate potentially damaging effects disruptions. this will be extremely challenging due many physics engineering constraints such as limitations port access amount species injected impurities. Additionally, questions relevant remain unsolved mechanisms for mixing assimilation impurities...
The ITER plasma control system has the same functional scope as systems in present tokamaks. These are operation scenario sequencing, basic (magnetic and kinetic), advanced (control of RWMs, NTMs, ELMs, error fields, etc) fast shutdown. This chapter considers only initiation control. describes progress achieved these areas tokamak experiments since Physics Basis (1999 Nucl. Fusion 39 2577) was written results assessment to provide done for design (15?MA machine) at a more detailed level than...
The first high-confinement mode (H-mode) with type-III edge localized modes at an H factor of IPB98( y ,2) ∼ 1 has been obtained about MW lower hybrid wave power on the EAST superconducting tokamak. H-mode plasma appeared after wall conditioning by lithium (Li) evaporation before breakdown and real-time injection fine Li powder into edge. threshold for access follows international tokamak scaling even in low density range a identified. With increasing accumulation deposited duration was...
DIII-D experiments on rapid shutdown runaway electron (RE) beams have improved the understanding of processes involved in RE beam control and dissipation. Improvements feedback enabled stable confinement out to volt-second limit ohmic coil, as well enabling a ramp down zero current. Spectroscopic studies shown that neutrals tend be excluded from centre. Measurements energy distribution function indicate broad with mean order several MeV peak energies 30–40 MeV. The appears more skewed...
Abstract Divertor detachment offers a promising solution to the challenge of plasma-wall interactions for steady-state operation fusion reactors. Here, we demonstrate excellent compatibility actively controlled full divertor with high-performance ( β N ~ 3, H 98 1.5) core plasma, using high-β p (poloidal beta, > 2) scenario characterized by sustained internal transport barrier (ITB) and modest edge (ETB) in DIII-D tokamak. The high- high-confinement facilitates which, turn, promotes...
The first suppression of the important and deleterious m = 2/n 1 neoclassical tearing mode (NTM) is reported using electron cyclotron current drive (ECCD) to replace 'missing' bootstrap in island O-point. Experiments on DIII-D tokamak verify that maximum shrinkage occurs when ECCD location coincides with q 2 surface. plasma control system put into a 'search suppress' make small changes toroidal field find lock onto optimum position, based real time measurements dBθ/dt, for complete NTM by...
High-pressure gas-jet injection of neon and argon is shown to be a simple robust method mitigate the deleterious effects disruptions on DIII-D tokamak. The gas jet penetrates central plasma at its sonic velocity. deposited species dissipates >95% by radiation substantially reduces mechanical stresses vessel caused poloidal halo currents. species-charge distribution can include >50% fraction neutral which inhibits runaway electrons. favorable scaling this technique burning fusion plasmas discussed.
Operating experimental devices have provided key inputs to the design process for ITER axisymmetric control. In particular, experiments quantified controllability and robustness requirements in presence of realistic noise disturbance environments, which are difficult or impossible characterize with modelling simulation alone. This kind information is particularly critical vertical control, poses highest demands on poloidal field system performance, since consequences loss control can be...
A severe consequence of a disruption on large tokamaks such as ITER could be the generation multi-megaelectronvolt electron beams that damage vacuum vessel and structures machine if they hit wall unmitigated. The mitigation runaway is thus key requirement for reliable operation ITER. In order to achieve mitigation, new fast shutdown technique has been developed: injection shattered cryogenic pellet in plasma, which expected increase density up levels where beam processes are mitigated by...
AbstractTo move to a fusion DEMO power plant after ITER, Fusion Nuclear Science Facility (FNSF) is needed in addition ITER and research operating tokamaks those under construction. The FNSF will enable on how utilize deal with the products of reactions, addressing such issues as extract energy from neutrons alpha particles into high-temperature process heat streams be either used directly or converted electricity, make tritium lithium, effects blanket structures, manage first wall surface...
In this paper, a linear model for plasma current, position and shape control based on the rigid motion assumption is presented implemented in an EAST tokamak simulator. The simulator models plasma, poloidal field (PF) coils, power supplies, used to verify algorithm optimize parameters PF coil current trajectories. Plasma has been achieved during last several operation campaigns due successful decoupling of shape. logic experimental results are described detail. Diverted shapes, including...
MHD simulations of rapid shutdown scenarios by massive particle injection in DIII-D, Alcator C-Mod and ITER are performed order to study runaway electron (RE) transport during mitigated disruptions. The include a RE confinement model using drift-orbit calculations for test particles. A comparison limited diverted plasma shapes is studied DIII-D simulations, improved the shape found due both spatial localization reduced toroidal spectrum nonlinear activity. compare which impurity (Ar)...
Key plasma physics and real-time control elements needed for robustly stable operation of high fusion power discharges in ITER have been demonstrated recent research worldwide.Recent analysis has identified the current density profile as main drive disruptive instabilities simulating ITER's baseline scenario with low external torque.Ongoing development model-based active magnetohydrodynamic is improving stability multiple scenarios.Significant advances made toward physicsbased prediction...
In a magnetic fusion reactor, the achievement of certain type plasma current profiles, which are compatible with magnetohydrodynamic stability at high pressure, is key to enable gain and non-inductive sustainment for steady-state operation. The approach taken toward establishing such profiles DIII-D tokamak create desired profile during ramp-up early flattop phases. evolution in time related poloidal flux, modeled normalized cylindrical coordinates using partial differential equation usually...
The requirements of the DIII-D physics program have led to development many operational control results with direct relevance ITER. These include new algorithms for robust and sustained stabilization neoclassical tearing modes electron cyclotron current drive, model-based controllers resistive wall mode in presence ELMs, coupled linear–nonlinear provide good dynamic axisymmetric while avoiding coil limits, adaptation plasma system (PCS) operate next-generation superconducting tokamaks....
New rapid shutdown strategies have been recently tested in the DIII-D tokamak to mitigate runaway electrons (REs). Disruptions ITER are predicted generate multi-MeV REs that could damage machine. The RE population large tokamaks is expected be dominated by avalanche amplification which can mitigated at high density levels collisional drag. Particle injection schemes for suppression of developed and ITER-relevant scenarios: massive gas injection, shattered pellet (SPI) shell injection....
In tokamak fusion plasmas, control of the spatial distribution profile toroidal plasma current plays an important role in realizing certain advanced operating scenarios. These scenarios, characterized by improved confinement, magnetohydrodynamic stability, and a high fraction non-inductively driven current, could enable steady-state reactor operation with gain. Current experiments at DIII-D focus on using combination feedforward feedback to achieve targeted during ramp-up early flat-top...
Research on the National Spherical Torus Experiment, NSTX, targets physics understanding needed for extrapolation to a steady-state ST Fusion Nuclear Science Facility, pilot plant, or DEMO. The unique operational space is leveraged test theories next-step tokamak operation, including ITER. Present research also examines implications coming device upgrade, NSTX-U. An energy confinement time, τE, scaling unified varied wall conditions exhibits strong improvement of BTτE with decreased electron...
Abstract A total power injection up to 0.3 GJ has been achieved in EAST long pulse H-mode operation of 101.2 s with an ITER-like water-cooled tungsten (W) mono-block divertor, which steady-state exhaust capability 10 MWm −2 . The peak temperature W target saturated at 12 the value T ~ 500 °C a heat flux ~3.3 MW m being maintained during discharge. By tailoring 3D divertor plasma footprint through edge magnetic topology change, load was broadly dispersed and thus sputtering were well...
Data on the discharge behavior, thermal loads, halo currents, and runaway electrons have been obtained in disruptions DIII-D tokamak [J. L. Luxon G. Davis, Fusion Technol. 8, 2A 441 (1985)]. These experiments also evaluated techniques to mitigate while minimizing electron production. Experiments injecting cryogenic impurity “killer” pellets of neon argon massive amounts helium gas successfully reduced these disruption effects. The current generation, scaling, mitigation are understood good...