A5045577349- Magnetic confinement fusion research
- Fusion materials and technologies
- Superconducting Materials and Applications
- Particle accelerators and beam dynamics
- Ionosphere and magnetosphere dynamics
- Plasma Diagnostics and Applications
- Laser-Plasma Interactions and Diagnostics
- Nuclear reactor physics and engineering
- Solar and Space Plasma Dynamics
- Nuclear Materials and Properties
- Semiconductor materials and devices
- Laser-induced spectroscopy and plasma
- Dust and Plasma Wave Phenomena
- Nuclear Physics and Applications
- Atomic and Molecular Physics
- Particle Accelerators and Free-Electron Lasers
- High-Energy Particle Collisions Research
- Metal and Thin Film Mechanics
- Diamond and Carbon-based Materials Research
- Silicon and Solar Cell Technologies
- Atomic and Subatomic Physics Research
- Spacecraft and Cryogenic Technologies
- Korean Peninsula Historical and Political Studies
- Meteorological Phenomena and Simulations
- Silicon Carbide Semiconductor Technologies
Princeton Plasma Physics Laboratory
2015-2024
Princeton University
2014-2023
University of Surrey
2020
Instituto de Ciencias Agrarias
2020
Hunan University
2018
Oak Ridge National Laboratory
2005-2014
General Atomics (United States)
1995-2013
Lawrence Livermore National Laboratory
1995-2013
Oak Ridge Associated Universities
1995-2010
Oak Ridge Institute for Science and Education
2010
A multi-machine database for the H-mode scrape-off layer power fall-off length, λq in JET, DIII-D, ASDEX Upgrade, C-Mod, NSTX and MAST has been assembled under auspices of International Tokamak Physics Activity. Regression inside finds that most important scaling parameter is poloidal magnetic field (or equivalently plasma current), with decreasing linearly increasing Bpol. For conventional aspect ratio tokamaks, regression , yielding λq,ITER ≅ 1 mm baseline inductive burning scenario at Ip...
Mastering nuclear fusion, which is an abundant, safe, and environmentally competitive energy, a great challenge for humanity. Tokamak represents one of the most promising paths toward controlled fusion. Obtaining high-performance, steady-state, long-pulse plasma regime remains critical issue. Recently, big breakthrough in steady-state operation was made on Experimental Advanced Superconducting (EAST). A with world-record pulse length 1056 s obtained, where density divertor peak heat flux...
Abstract The objectives of NSTX-U research are to reinforce the advantages STs while addressing challenges. To extend confinement physics low- A , high beta plasmas lower collisionality levels, understanding transport mechanisms that set performance and pedestal profiles is being advanced through gyrokinetic simulations, reduced model development, comparison NSTX experiment, as well improved simulation RF heating. develop stable non-inductive scenarios needed for steady-state operation,...
The National Spherical Torus Experiment (NSTX) is being built at Princeton Plasma Physics Laboratory to test the fusion physics principles for spherical torus concept MA level. NSTX nominal plasma parameters are R0 = 85 cm, a 67 R/a ⩾ 1.26, Bt 3 kG, Ip 1 MA, q95 14, elongation κ ⩽ 2.2, triangularity δ 0.5 and pulse length of up 5 s. heating/current drive tools high harmonic fast wave (6 MW, s), neutral beam injection (5 80 keV, s) coaxial helicity injection. Theoretical calculations predict...
The two-dimensional radial vs poloidal structure and motion of edge turbulence in the National Spherical Torus Experiment (NSTX) were measured using high-speed imaging visible light emission from a localized neutral gas puff. Edge images are shown analysed for Ohmic, L- H-mode plasma conditions. often show regions strong known as 'blobs', which move both poloidally radially at typical speed ≈105 cm s−1, sometimes spatially periodic features.
The pressure at the top of edge transport barrier (or ‘pedestal height’) strongly impacts fusion performance, while large localized modes (ELMs), driven by free energy in pedestal region, can constrain material lifetimes. Accurately predicting height and ELM behavior ITER is an essential element prediction optimization performance. Investigation intermediate wavelength MHD ‘peeling–ballooning’ modes) has led to improved understanding important constraints on mechanism for ELMs. combination...
Abstract The spherical tokamak (ST) is a leading candidate for Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. National Spherical Torus eXperiment (NSTX) MA-class ST facility in the US actively developing physics basis an ST-based FNSF. In plasma transport research, experiments exhibit strong (nearly inverse) scaling of normalized confinement with collisionality, if this trend holds at low high fusion neutron fluences could be achievable very...
Reduction or elimination of edge localized modes (ELMs) while maintaining high confinement is essential for future fusion devices, e.g., the ITER. An ELM-free regime was recently obtained in National Spherical Torus Experiment, following lithium (Li) evaporation onto plasma-facing components. Edge stability calculations indicate that pre-Li discharges were unstable to low-$n$ peeling ballooning modes, broader pressure profiles stabilized post-Li discharges. Normalized energy increased by 50%...
National Spherical Torus Experiment [which M. Ono et al., Nucl. Fusion 40, 557 (2000)] high-power divertor plasma experiments have shown, for the first time, that benefits from lithium coatings applied to facing components found previously in limited plasmas can occur also diverted configurations. Lithium were with pellets injected into helium discharges, and an oven directed a collimated stream of vapor toward graphite tiles lower center stack divertor. depositions few milligrams 1g been...
The XGC1 edge gyrokinetic code is used for a high fidelity prediction the width of heat-flux to divertor plates in attached plasma condition. simulation results are validated against empirical scaling $\lambda_q \propto B_P^{-\gamma}$ obtained from present tokamak devices, where $\lambda_q$ mapped outboard midplane and $\gamma_q=1.19$ as defined by T. Eich et al. [Nucl. Fusion 53 (2013) 093031], $B_P$ magnitude poloidal magnetic field at separatrix surface. This predicts \leq 1mm$ when...
Periods of edge localized mode (ELM)-free H-mode with increased pedestal pressure and width were observed in the DIII-D tokamak when density fluctuations to region near separatrix present. Injection a powder 45 µm diameter lithium particles duration enhanced phases up 350 ms, also likelihood transition phase. Lithium injection at level sufficient for triggering extended resulted significant plasma core, but carbon other higher Z impurities as well radiated power levels reduced. Recycling...
Abstract A fusion nuclear science facility (FNSF) could play an important role in the development of energy by providing environment needed to develop materials and components. The spherical torus/tokamak (ST) is a leading candidate for FNSF due its potentially high neutron wall loading modular configuration. key consideration choice configuration range achievable missions as function device size. Possible include: fluence, demonstrating tritium self-sufficiency, electrical self-sufficiency....
A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out 2010 order construct a predictive scaling relation applicable next step devices including ITER, FNSF, DEMO. Few published laws are available those that have been were obtained under widely varying conditions geometries, leading conflicting predictions for this critically important quantity. This study designed overcome these...
Abstract Lithium wall coatings have been shown to reduce recycling, suppress edge-localized modes (ELMs), and improve energy confinement in the National Spherical Torus Experiment (NSTX). Here we document effect of gradually increasing lithium on discharge characteristics, with reference ELMy discharges obtained boronized, i.e. non-lithiated conditions. We observed a continuous but not quite monotonic reduction recycling improvement confinement, gradual alteration edge plasma profiles,...
A critical challenge facing the basic long-pulse high-confinement operation scenario ($H$ mode) for ITER is to control a magnetohydrodynamic (MHD) instability, known as edge localized mode (ELM), which leads cyclical high peak heat and particle fluxes at plasma components. breakthrough made in Experimental Advanced Superconducting Tokamak achieving new steady-state $H$ without presence of ELMs duration exceeding hundreds energy confinement times, by using novel technique continuous real-time...
Abstract Ion temperatures of over 100 million degrees Kelvin (8.6 keV) have been produced in the ST40 compact high-field spherical tokamak (ST). excess 5 keV not previously reached any ST and only obtained much larger devices with substantially more plasma heating power. The corresponding fusion triple product is calculated to be <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mrow> <mml:msub> <mml:mi>n</mml:mi> <mml:mi>i</mml:mi> <mml:mn>0</mml:mn>...
Research in NSTX has been conducted to establish spherical torus plasmas be used for high β, auxiliary heated experiments. This device a major radius R0 = 0.86 m and midplane halfwidth of 0.7 m. It operated with toroidal magnetic field B0 ⩽ 0.3 T Ip 1.0 MA. The evolution the plasma equilibrium is analysed between discharges an automated version EFIT code. Limiter, double null lower single diverted configurations have sustained several energy confinement times. stored reached 92 kJ (βt 17.8%)...
The National Spherical Torus Experiment (NSTX) has demonstrated the advantages of low aspect ratio geometry in accessing high toroidal and normalized plasma beta, βN ≡ 108⟨βt⟩ aB0/Ip. Experiments have reached βt = 39% 7.2 through boundary profile optimization. High plasmas can exceed ideal no-wall stability limit, βNno−wall, for periods much greater than wall eddy current decay time. Resistive mode (RWM) physics is studied to understand stabilization these plasmas. spectrum unstable RWMs...
Extensive lithium wall coatings and liquid plasma-limiting surfaces reduce recycling, with dramatic improvements in Ohmic plasma discharges the Current Drive Experiment-Upgrade. Global energy confinement times increase by up to 6 times. These results exceed scalings such as $\mathrm{ITER}98\mathrm{P}(y,1)$ $2--3$ times, represent largest ever observed for an tokamak plasma. Measurements of ${D}_{\ensuremath{\alpha}}$ emission indicate that global recycling coefficients decrease approximately...
Fusion power has been increased by a factor of 3 in DIII-D tailoring the pressure profile to avoid kink instability $H$-mode plasmas. The resulting plasmas are found have neoclassical ion confinement. This reduction transport losses beam-heated with negative central shear is correlated dramatic density fluctuations. Improved magnetohydrodynamic stability achieved controlling plasma width. In deuterium highest gain $Q$ (the ratio fusion input power), was 0.0015, corresponding an equivalent...
Experiments in the National Spherical Torus Experiment (NSTX) have shown beneficial effects on performance of divertor plasmas as a result applying lithium coatings graphite and carbon-fiber-composite plasma-facing components. These mostly been applied by pair evaporators mounted at top vacuum vessel which inject collimated streams vapor toward lower divertor. In neutral beam injection (NBI)-heated deuterium H-mode run immediately after application lithium, modifications included decreases...
An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore attractive configurations and define key research development (R&D) areas. The large size mass predicted by earlier plant studies had led cost projections much higher than those the advanced tokamak plant. As such, first major goal ARIES-CS was investigate if plants can be made comparable in variants while maintaining desirable properties. fusion core components would have complex shapes...
The NSTX operates at low aspect ratio (R/a ∼ 1.3) and high beta (up to 40%), allowing tests of global confinement local transport properties that have been established from higher devices. plasmas are heated by up 7 MW deuterium neutral beams with preferential electron heating as expected for ITER. Confinement scaling studies indicate a strong BT dependence, current dependence is weaker than observed ratio. Dimensionless experiments increase in decreasing collisionality weak degradation...
Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation economically viable fusion power reactors. To date, few demonstrations exist this approach in diverted tokamak and we here provide an overview such work on National Spherical Torus Experiment (NSTX). The Lithium Divertor (LLD) was installed operated for 2010 run campaign using evaporated coatings filling method. LLD consisted copper-backed structure with porous...