A5025852561- 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
- Solar and Space Plasma Dynamics
- Nuclear Physics and Applications
- Laser-induced spectroscopy and plasma
- Cold Fusion and Nuclear Reactions
- Nuclear Engineering Thermal-Hydraulics
- Natural Language Processing Techniques
- Atomic and Subatomic Physics Research
- Quantum, superfluid, helium dynamics
- Rocket and propulsion systems research
- Dust and Plasma Wave Phenomena
- Parallel Computing and Optimization Techniques
- Laser Design and Applications
- Tunneling and Rock Mechanics
- Tribology and Lubrication Engineering
- Silicon Carbide Semiconductor Technologies
- Spacecraft and Cryogenic Technologies
- Korean Peninsula Historical and Political Studies
University of Washington
2015-2024
American Institute of Aeronautics and Astronautics
2011-2024
Seattle University
1993-2023
Kyushu University
2019
Princeton Plasma Physics Laboratory
2001-2017
Office of Science
2016
Lawrence Livermore National Laboratory
2001-2014
Princeton University
2003-2014
University of Wisconsin–Madison
2013-2014
Oak Ridge National Laboratory
2003-2007
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...
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...
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...
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....
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,...
Dedicated experiments in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion, 42, 614 (2002)], Joint European Torus (JET) [P. H. Rebut, R. J. Bickerton, and B. E. Keen, Fusion 25, 1011 (1985)], National Spherical Experiment (NSTX) [M. Ono, S. M. Kaye, Y.-K. Peng et al., 40, 557 (2000)] reveal commonalities of resistive wall mode (RWM) stabilization by sufficiently fast toroidal plasma rotation devices different size aspect ratio. In each device weakly damped n=1 RWM manifests itself resonant field...
We report the first results of nondisruptive, central fueling a tokamak by injection an accelerated spheromak compact toroid (CT). Interferometry measurements indicate plasma on fast time scale (0.5 ms), with more than 50% CT mass used for fueling. The particle inventory increased 30% without disruption.
Steady-state handling of divertor heat flux is a critical issue for ITER and future conventional spherical tokamaks with compact high-power density divertors. A novel ‘snowflake’ (SFD) configuration was theoretically predicted to have significant magnetic geometry benefits mitigation, such as an increased plasma-wetted area higher volume available volumetric power momentum loss processes, compared the standard divertor. Both peak reduction impurity screening been achieved simultaneously core...
Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of heat flux and plate erosion remains to critical issues ITER future concept devices based on conventional spherical geometry with high power density divertors. Experiments conducted in 4–6 MW NBI-heated H-mode plasmas demonstrated is compatible high-confinement core plasma...
The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in summer of 2015.NSTX-U first plasma subsequently achieved, diagnostic control systems have been commissioned, H-mode accessed, magnetic error fields identified mitigated, physics research campaign carried out.During ten run weeks operation, NSTX-U surpassed record pulse-durations toroidal (TF), high-performance ~1 MA plasmas comparable to best sustained near...
The formation of an elongated Sweet-Parker current sheet and a transition to plasmoid instability has for the first time been predicted by simulations in large-scale toroidal fusion plasma absence any preexisting instability. Plasmoid is demonstrated through resistive MHD transient coaxial helicity injection experiments National Spherical Torus Experiment (NSTX). Consistent with theory, fundamental characteristics instability, including fast reconnection rate, have observed these realistic...
Coaxial Helicity Injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240kA of toroidal current without use central solenoid.Values multiplication ratio (CHI / injector current) up to 10 were obtained, in agreement with predictions.The discharges which lasted for 200ms, limited only by programmed waveform are more than an order magnitude longer duration that any CHI previously a Spheromak or (ST).
H-mode operation plays a crucial role in National Spherical Torus Experiment (NSTX) research, allowing higher beta limits due to reduced plasma pressure peaking, and long pulse high bootstrap current fraction. Here, new results are presented the areas of edge localized modes (ELMs), pedestal physics power threshold studies. ELMs several types as reported by aspect ratio tokamaks have been observed: (1) large, Type I ELMs, (2) intermediate-sized III (3) tiny ELMs. Many performance discharges...
The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order advance spherical torus (ST) concept. In do this, NSTX utilizes up 7.5 MW neutral beam injection, 6 harmonic fast waves (HHFWs), it operates with plasma currents 1.5 MA elongations 2.6 a field 0.45 T. New facility, diagnostic modelling capabilities developed over past two years have enabled research team make significant progress...
Transient coaxial helicity injection (CHI) started discharges in the National Spherical Torus Experiment (NSTX) have attained peak currents up to 300 kA and when coupled induction, it has produced 200 additional current over inductive-only operation. CHI NSTX shown be energetically quite efficient, producing a plasma of about 10 A/J capacitor bank energy. In addition, for first time, CHI-produced toroidal that couples induction continues increase with energy supplied by power supply at...
Developing a reactor-compatible divertor has been identified as particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted improved H-mode confinement, power threshold reduction, and other plasma performance benefits. During the 2010 campaign, application relatively modest amount Li (300 mg prior to discharge) ∼50% reduction heat load on liquid (LLD) attributable enhanced bolometric radiation. These promising results related...
The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. torus (ST) confinement concept projected to operate at high toroidal beta a fraction the non-inductive bootstrap current as required for efficient reactor system. use conventional in ST-based fusion nuclear facility generally believed not be possibility. Solenoid-free plasma start-up therefore area extensive worldwide research activity. also relevant steady-state tokamak operation, transformer...
Simulations of massive gas injection for disruption mitigation in DIII-D are carried out to compare the toroidal peaking radiated power cases one and two jets. The radiation factor (TPF) results from a combination distribution impurities heat flux associated with mode. When ignoring effects strong uni-directional neutral beam rotation present experiment, injected found spread helically along field lines preferentially toward high-field-side, which is explained terms nozzle equation....
The first successful results on the transfer of a coaxial helicity injection (CHI) produced discharge to inductive operation are reported. CHI-assisted plasma startup is more robust than only operation. After hand off for operation, initial 90 kA CHI-produced current drops 40 kA, then ramps up 170 using 30 mV s, 30% higher that by induction alone. These significant performance enhancing were obtained HIT-II spherical torus experiment (major/minor radius 0.3/0.2 m).
We report the observation of a high performance scenario in National Spherical Torus Experiment with very small edge-localized modes (ELMs). These ELMs, individually, have no measurable impact on stored energy and are observed by several diagnostics. The ELMs clear differences as compared ELM types reported literature, this operating mode has distinct features other tokamak scenarios little or ELMs. is termed 'type V', it short-lived n = 1 magnetic precursor oscillation rotating counter to...