R. Majeski
A5022400747- Magnetic confinement fusion research
- Particle accelerators and beam dynamics
- Fusion materials and technologies
- Superconducting Materials and Applications
- Plasma Diagnostics and Applications
- Ionosphere and magnetosphere dynamics
- Laser-Plasma Interactions and Diagnostics
- Particle Accelerators and Free-Electron Lasers
- Gyrotron and Vacuum Electronics Research
- Atomic and Molecular Physics
- Solar and Space Plasma Dynamics
- Atomic and Subatomic Physics Research
- Nuclear Materials and Properties
- Nuclear reactor physics and engineering
- Nuclear Physics and Applications
- Laser-induced spectroscopy and plasma
- Semiconductor materials and devices
- Korean Peninsula Historical and Political Studies
- Cold Atom Physics and Bose-Einstein Condensates
- Muon and positron interactions and applications
- Dust and Plasma Wave Phenomena
- Acoustic Wave Resonator Technologies
- Spacecraft and Cryogenic Technologies
- Astro and Planetary Science
- High voltage insulation and dielectric phenomena
Princeton Plasma Physics Laboratory
2015-2024
Princeton University
2005-2023
Oak Ridge National Laboratory
1993-2020
University of Washington
2001-2018
University of California, Los Angeles
2018
American Institute of Aeronautics and Astronautics
2018
Lawrence Livermore National Laboratory
2001-2013
University of California, San Diego
2009
University of Wisconsin System
1992-2003
Northrop Grumman (United States)
2002
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...
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...
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...
Peak fusion power production of 6.2\ifmmode\pm\else\textpm\fi{}0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total 29.5 MW. These have an inferred central alpha particle density 1.2\ifmmode\times\else\texttimes\fi{}${10}^{17}$ ${\mathrm{m}}^{\mathrm{\ensuremath{-}}3}$ without the appearance either disruptive magnetohydrodynamics events or detectable changes Alfv\'en wave activity. The measured loss rate energetic particles agreed with...
We report the first observation of global recycling coefficient R near 0.5 in Lithium Tokamak eXperiment-β (LTX-β), significantly below minimum previously reported other devices. In a series experiments with varied Li wall conditioning, estimates have been made using Lyman-α array and DEGAS2 modeling. A progressive reduction emission increased lithium an increase edge electron temperature are observed. It is also observed that increasing coating thickness, effective particle confinement time...
Alpha-particle-driven toroidal Alfvén eigenmodes (TAEs) have been observed for the first time in deuterium-tritium (D-T) plasmas on tokamak fusion test reactor (TFTR). These modes are 100–200 ms following end of neutral beam injection with reduced central magnetic shear and elevated safety factor [q0>1]. Mode activity is localized to region discharge r/a<0.5 fluctuation level B̃⊥/B∥∼10−5 mode numbers range n=2–4, consistent theoretical calculations α-TAE stability TFTR.Received 11 November...
Recent experiments in the Current Drive Experiment-Upgrade (CDX-U) provide a first-ever test of large area liquid lithium surfaces as tokamak first wall to gain engineering experience with metal and investigate whether very low recycling plasma regimes can be accessed walls. The CDX-U is compact (R = 34 cm, 22 Btoroidal 2 kG, IP 100 kA, Te(0)∼ eV, ne(0) ∼ 5 × 1019 m−3) spherical torus at Princeton Plasma Physics Laboratory. A toroidal pool limiter an 2000 cm2 (half total limiting surface)...
After many years of fusion research, the conditions needed for a D–T reactor have been approached on Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For first time unique phenomena present in plasma are now being studied laboratory plasma. The magnetic experiments to study plasmas using nearly equal concentrations deuterium and tritium carried out TFTR. At maximum power 10.7 MW, 39.5 MW neutral-beam heating, supershot discharge 6.7 high-βp following current rampdown....
Experimental and theoretical results pertaining to the collisional drift instability of a weakly ionized plasma are presented. A non-local cylindrical model is developed solved numerically yield dispersion characteristics which compared with local slab relation. The shown be inadequate in describing how mode stability varies azimuthal number. Detailed experimental observations single coherent oscillations argon also identified as waves on basis comparison theory. Moreover, demonstrate that...
It has been predicted for over a decade that low-recycling plasma-facing components in fusion devices would allow high edge temperatures and flat or nearly temperature profiles. In recent experiments with lithium wall coatings the Lithium Tokamak Experiment (LTX), hot ($>200\text{ }\text{ }\mathrm{eV}$) electron profiles have measured following termination of external fueling. Reduced recycling was demonstrated by retention $\ensuremath{\sim}60%$ injected hydrogen walls discharge. Electron...
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...
The Tokamak Fusion Test Reactor (TFTR) (R. J. Hawryluk, to be published in Rev. Mod. Phys.) experiments on high-temperature plasmas, that culminated the study of deuterium–tritium D–T plasmas containing significant populations energetic alpha particles, spanned over two decades from conception completion. During design TFTR, key physics issues were magnetohydrodynamic (MHD) equilibrium and stability, plasma energy transport, impurity effects, reactivity. Energetic particle was given less...
AbstractThe mission of the National Spherical Torus Experiment (NSTX) is to prove principles spherical torus physics by producing high-βt plasmas that are noninductively sustained and whose current profiles in steady state. The NSTX will be one first ultralow-aspect-ratio tori (R/a ≤ 1.3) operate at high power (Pinput up 11 MW) produce (25 40%), low-collisionality, high-bootstrap-fraction (≤70%) discharges. Both radio-frequency neutral beam heating drive employed. Built into sufficient...
The Tomamak Fusion Test reactor has performed initial high-power experiments with the plasma fueled nominally equal densities of deuterium and tritium. Compared to pure plasmas, energy stored in electron ions increased by \ensuremath{\sim}20%. These increases indicate improvements confinement associated use tritium possibly heating electrons \ensuremath{\alpha} particles created D-T fusion reactions.
Liquid metal (LM) plasma-facing components (PFCs) may provide a resolution to the challenging fusion environment, particularly first wall and divertor surfaces. Transforming these concepts into viable technologies will require considerable research development. With nuclear regime in mind, Fusion Energy System Studies group examined LM PFCs order identify needed thrusts that could accelerate their development assess viability. behavior, solid substrate aspects, facility integration aspects...
Ion cyclotron emission (ICE) has been observed during D-T discharges in the Tokamak Fusion Test Reactor (TFTR), using RF probes located near top and bottom of vacuum vessel. Harmonics alpha frequency ( Omega a,) evaluated at outer midplane plasma edge are onset beam injection phase TFTR supershots persist for approximately 100-250 ms. These results contrast with observations ICE JET, which harmonics a evolve population edge. Such differences believed to be due fact that newly born fusion...
An approach to obtaining efficient single-pass-mode conversion at high parallel wave number from the fast magnetosonic slow ion Bernstein wave, in a two-ion species tokamak plasma, is described. The intent produce localized electron heating or current drive via mode-converted wave. In particular, this technique can be adapted off-axis for profile control. Modeling case of deuterium-tritium plasmas TFTR presented.
The Current Drive Experiment-Upgrade [T. Munsat, P. C. Efthimion, B. Jones, R. Kaita, Majeski, D. Stutman, and G. Taylor, Phys. Plasmas 9, 480 (2002)] spherical tokamak research program has focused on lithium as a large area plasma-facing component (PFC). energy confinement times showed sixfold or more improvement over discharges without PFCs. This was an increase of up to factor 3 ITER98P(y,1) scaling [ITER Physics Basis Editors, Nucl. Fusion 39, 2137 (1999)], reflects the largest...
Use of a large-area liquid lithium limiter in the CDX-U tokamak produced largest relative increase (an enhancement factor 5–10) Ohmic confinement ever observed. The results from do not agree with existing scaling laws, and cannot easily be projected to new experiment (LTX). Numerical simulations low recycling discharges have now been performed ASTRA-ESC code special reference transport model suitable for diffusion-based regime, incorporating boundary conditions nonrecycling walls, fuelling...
AbstractThe use of a fusion component testing facility to study and establish, during the ITER era, remaining scientific technical knowledge needed by Demo is considered described in this paper. This aims test components an integrated nuclear environment, for first time, discover understand underpinning physical properties, develop improved further testing, time-efficient manner. It requires design with extensive modularization remote handling activated components, flexible hot-cell...
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...
The first-ever successful operation of a tokamak with large area (40% the total plasma surface area) liquid lithium wall has been achieved in Lithium Tokamak eXperiment (LTX). These results were obtained new, electron beam-based evaporation system, which can deposit coating on limiting LTX five-minute period. Preliminary analyses diamagnetic and other data for discharges operated indicate that confinement times increased by 10× compared to helium-dispersed solid coatings. Ohmic energy fresh...