A5068985781- 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
- Nuclear reactor physics and engineering
- Advanced Data Storage Technologies
- Frequency Control in Power Systems
- Physics of Superconductivity and Magnetism
- Atomic and Subatomic Physics Research
- Magnetic Field Sensors Techniques
- Electrostatic Discharge in Electronics
- Nuclear Physics and Applications
- Laser-Plasma Interactions and Diagnostics
- Nuclear Materials and Properties
- CCD and CMOS Imaging Sensors
- Geotechnical and Geomechanical Engineering
- Gas Dynamics and Kinetic Theory
- Laser-induced spectroscopy and plasma
- Solar and Space Plasma Dynamics
- Electromagnetic Launch and Propulsion Technology
- Copper Interconnects and Reliability
- Plasma and Flow Control in Aerodynamics
- Radiation Effects in Electronics
General Atomics (United States)
2012-2024
DIII-D National Fusion Facility
2008-2013
Oak Ridge National Laboratory
2007
Los Alamos National Laboratory
2003
Lockheed Martin Orincon (United States)
1997
National Grid (United States)
1971
A practical method for performing a tokamak equilibrium reconstruction in real time arbitrary varying discharge shapes and current profiles is described. An approximate solution to the Grad-Shafranov relation found which best fits diagnostic measurements. Thus, spatial distribution of poloidal flux toroidal density available that consistent with plasma force balance, allowing accurate evaluation parameters such as shape safety factor profile. The solutions are produced at rate sufficient...
New experiments on JET, COMPASS-D and DIII-D have identified the critical scalings of error field sensitivity harmonic content effects, enabling predictions to be made requirements for larger devices such as ITER. Thresholds are lowest at low density, a regime proposed H mode access Results suggest moderate (δB/B ≈ 10-4) ITER, comparable with size its intrinsic error, although there uncertainties in scaling behaviour. Other studies show that sideband harmonics (2, 1) component play an...
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...
This paper presents the results of a multi-system codes benchmarking study recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy (TESC), using different methodologies to arrive at CFETR performance parameters under same constraints show that correlation between physics fusion is consistent, computed are in good agreement. Optimization first...
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...
Abstract Physics-based simulations project a compact net electric fusion pilot plant with nuclear testing mission is possible at modest scale based on the advanced tokamak concept, and identify key parameters for its optimization. These utilize new integrated 1.5D core-edge approach whole device modeling to predict performance by self-consistently applying transport, pedestal current drive models converge fully non-inductive stationary solutions, predicting profiles energy confinement given...
An artificial neural network, combining signals from a large number of plasma diagnostics, was used to estimate the high- beta disruption boundary in DIII-D tokamak. It is shown that inclusion many diagnostic measurements results much more accurate prediction than provided by traditional Troyon limit. A trained network constitutes non-linear, non-parametric model boundary. Through analysis input-output sensitivities, relative statistical significance various (plasma parameters) for...
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....
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...
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...
The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, current, is a critical element for achieving mission performance. Analysis PF capabilities has focused the 15 MA Q = 10 scenario with 300–500 s flattop burn phase. operating space available ELMy H-mode discharges in ITER upgrades to coils or associated systems establish confidence that objectives can be reached have been identified. Time dependent self-consistent...
A comprehensive systems code that includes a range of physics and engineering considerations along with simplified costing model has been utilized to evaluate the primary cost drivers for compact tokamak pilot plant. The benchmarked against several reactor designs is sophisticated optimization algorithms develop optimal solutions set user-specified assumptions design constraints. In contrast previous models have focused on electricity as key metric, this study uses estimated capital...
Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks, leading confinement and abrupt termination plasma current (disruptions). The best examples the locked mode resistive wall mode. Understanding underlying field anomalies (departures hardware-related from ideal toroidal poloidal on a single axis) interaction with them is crucial tokamak development. Results both experiments (Scoville J.T. La Haye R.J. 2003 Nucl. Fusion 43 250) (Garofalo A.M.,...
Experiments on the stabilization of highly elongated, vertically unstable plasmas were carried out DIII-D tokamak. Identification closed-loop transfer function showed that vertical stability could be usefully modelled as a second order dynamical system. The effect varying controller gains and field decay index was studied found to qualitatively predicted by low model proposed previously. implementation new hybrid inboard/outboard coil positional control with differing dynamics allowed...
AbstractThe ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Three are currently planned for the program. The ARIES-I design DT-burning reactor based on "modest" extrapolations from present physics database relies either existing technology or which trends already in place, often programs outside fusion; ARIES-II will employ potential advances physics; ARIES-III conceptual D-3He...
The objective of the Fusion Development Facility (FDF) under consideration is to carry forward advanced tokamak physics for optimization fusion reactors and enable development fusion's energy applications. A concept FDF based on approach with conservative expressions nonsuperconducting magnet technology presented. It envisioned nominally provide 2 MW/m2 neutron wall loading operate continuously up weeks as required nuclear component research development. will have tritium breeding capability...
The mission of the National Spherical Torus Experiment (NSTX) is demonstration physics basis required to extrapolate next steps for spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based (ST-CTF), and support ITER. Key issues are transport, steady state high β operation. To better understand electron new high- k scattering diagnostic was used extensively investigate gyro-scale fluctuations with varying temperature gradient scale length. Results from n = 3...
Understanding the dynamics of plasma startup and termination is important for present tokamaks predictive modeling future burning devices such as ITER. We report on experiments in DIII-D tokamak that explore rampdown phases benchmarking transport models. Key issues have been examined initiation burnthrough with limited inductive voltage achieving flattop maximum burn within technical limits coil systems their actuators while maintaining desired q profile. Successful requires scenarios...
This paper reports upon the development of new multi-area dispatch algorithms, their implementation and use in system operations.
Along with an expanded evaluation of the equilibrium operating space Korea Superconducting Tokamak Advanced Research, KSTAR, experimental equilibria most recent plasma discharges were reconstructed using EFIT code. In near-circular plasmas created in 2009, reached a stored energy 54 kJ maximum current 0.34 MA. Highly shaped near double-null configuration 2010 achieved H-mode clear edge localized mode (ELM) activity, and transiently up to 257 kJ, elongation 1.96 normalized beta 1.3. The 0.7...