A5034141457- Magnetic confinement fusion research
- Superconducting Materials and Applications
- Fusion materials and technologies
- Particle accelerators and beam dynamics
- Ionosphere and magnetosphere dynamics
- Nuclear reactor physics and engineering
- Laser-Plasma Interactions and Diagnostics
- Plasma Diagnostics and Applications
- Nuclear Materials and Properties
- Nuclear Engineering Thermal-Hydraulics
- Solar and Space Plasma Dynamics
- Nuclear Physics and Applications
- Cold Fusion and Nuclear Reactions
- Atomic and Subatomic Physics Research
- Physics of Superconductivity and Magnetism
- Electromagnetic Launch and Propulsion Technology
- Spacecraft Design and Technology
- Particle physics theoretical and experimental studies
- Frequency Control in Power Systems
- Metallurgy and Material Forming
- Computational Physics and Python Applications
- Astronomical Observations and Instrumentation
- Metallurgical Processes and Thermodynamics
- Engineering Applied Research
- Electric Motor Design and Analysis
Oak Ridge National Laboratory
1992-2024
Princeton Plasma Physics Laboratory
2010-2019
Princeton University
2008-2019
Lawrence Livermore National Laboratory
2006-2013
Lehigh University
2011
Tech-X Corporation (United States)
2011
Higashihiroshima Medical Center
2006
Hiroshima University
2006
University of California, Irvine
2006
University of Washington
2006
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...
A tokamak plasma configuration is reported that simultaneously improves on the maximum stable pressure, bootstrap current contribution, and kinetic stability to temperature density gradient driven modes in toroidal geometry. It characterized by negative magnetic shear interior a peaked pressure profile. Stability ideal low-n external kink requires conducting shell at 1.3 times minor radius. This novel promising for improved performance advanced experiments.
Significant progress has been made in the area of advanced modes operation that are candidates for achieving steady state conditions a fusion reactor. The corresponding parameters, domain operation, scenarios and integration issues discussed this chapter. A review presently developed scenarios, including discussions on operational space, is given. heating current drive recent years, especially off-axis drive, which essential achievement required profile. actuators necessary to produce...
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...
As part of the ITER Design Review and in response to issues identified by Science Technology Advisory Committee, physics requirements were reviewed as appropriate updated. The focus this paper will be on recent work affecting design with special emphasis topics near-term procurement arrangements. This describe results on: sensitivity studies, poloidal field coil requirements, vertical stability, effect toroidal ripple thermal confinement, material choice heat load for plasma-facing...
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...
The ITER divertor design is the culmination of years physics and engineering effort, building confidence that this critical component will satisfy requirements meet challenge burning plasma operation. With 54 cassette assemblies, each weighing ∼9 tonnes, nearly 3900 actively cooled high heat flux elements rated to steady-state surface power densities 10 MW m−2 a total ∼60 000 carbon fibre composite monoblocks ∼260 tungsten monoblocks/flat tiles, be largest most advanced its kind ever...
A Korean fusion energy development promotion law (FEDPL) was enacted in 2007. As a following step, conceptual design study for steady-state demonstration reactor (K-DEMO) initiated 2012. After the thorough 0D system analysis, parameters of main machine characterized by major and minor radii 6.8 2.1 m, respectively, were chosen further study. The analyses heating current drives performed plasma operation scenarios. Preliminary results on lower hybrid neutral beam drive are included herein....
The object of this review is to summarize the achievements research on Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] place that in context quest for practical fusion energy. a compact, high-field tokamak, whose unique design operating parameters have produced wealth new important results since it began operation 1993, contributing data extends tests critical physical models into parameter ranges regimes. Using only...
Compact optimized stellarators offer novel solutions for confining high-β plasmas and developing magnetic confinement fusion. The three-dimensional plasma shape can be designed to enhance the magnetohydrodynamic (MHD) stability without feedback or nearby conducting structures provide drift-orbit similar tokamaks. These configurations possibility of combining steady-state low-recirculating power, external control, disruption resilience previous with low aspect ratio, high β limit, good...
Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd Plasma Phys. Control. 46 B477) on the Spherical Tokamak (or Torus, ST) (Peng 2000 Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of conditions engineering a Component Test Facility (CTF), (Cheng 1998 Eng. Des. 38 219) which is very valuable step in development practical fusion energy....
A 20 MW/5 GHz lower hybrid current drive (LHCD) system was initially due to be commissioned and used for the second mission of ITER, i.e. Q = 5 steady state target. Though not part currently planned procurement phase, it is now under consideration an earlier delivery. In this paper, both physics technology conceptual designs are reviewed. Furthermore, appropriate work plan also developed. This design, R&D, installation a MW LHCD on ITER follows Scientific Technical Advisory Committee...
A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. device ultimately capable of small net electricity production in as compact facility possible and configuration scalable to full-size power plant. key capability for pilot-plant programme the high neutron fluence enabling nuclear science technology (FNST) research. It found that physics assumptions between those assumed ITER nth-of-a-kind it provide FNST-relevant wall loading devices. Thus, may be...
Sustainment of Q ∼ 10 operation with a fusion power ∼500 MW for several hundred seconds is key mission goal the ITER Project. Past calculations and simulations predict that these conditions can be produced in high-confinement mode (H-mode) at 15 MA relying on only inductive current drive. Earlier development baseline plasma scenarios provided focal point Design Review conducted 2007–2008. In intervening period, detailed predictive simulations, supported by experimental demonstrations...
The MIT Plasma Science and Fusion Center collaborators are proposing a high-performance Advanced Divertor RF tokamak eXperiment (ADX)-a specifically designed to address critical gaps in the world fusion research programme on pathway next-step devices: nuclear science facility (FNSF), pilot plant (FPP) and/or demonstration power (DEMO). This high-field (>= 6.5 T, 1.5 MA), high density (P/S similar MW m(-2)) will test innovative divertor ideas, including an 'X-point target divertor' concept,...
The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter complex fusion nuclear regime, and its technical mission attributes are being developed. FNSF one part of energy development pathway commercial power plant with other major components pre-FNSF research development, in parallel FNSF, pre-DEMO demonstration (DEMO). Energy Systems Studies group is developing basis for order provide a better understanding demands on plasma science programs.
Abstract The SOLPS-ITER code is utilized to analyze the boundary plasma associated with a fast-flow lithium (Li) divertor configuration in fusion nuclear science facility (FNSF) tokamak and identify operational regimes acceptable core conditions. Plasma transport from has been coupled liquid metal (LM) MHD/heat transfer model Li open-surface design assess its impact on scrape-off-layer (SOL) performance. Simulations only Neon (Ne) impurity seeding have conducted evaluate meeting FNSF demands...
The ideal magnetohydrodynamic (MHD) stability limits of low aspect ratio tokamak plasmas are computed numerically for with a range cylindrical safety factors q*, normalized plasma pressures beta , elongations kappa and central q(0). Four distinct regimes optimized, namely: (a) low-q* q(0)=1.1 without stabilizing wall, (b) no wall 1.1<q(0)<2, (c) high- high bootstrap fraction at moderate requiring edge current drive (d) very to but little external drive. A stable equilibrium is found an A=1.4...
Integrated simulations are performed to establish a physics basis, in conjunction with present tokamak experiments, for the operating modes International Thermonuclear Experimental Reactor (ITER). Simulations of hybrid mode done using both fixed and free-boundary 1.5D transport evolution codes including CRONOS, ONETWO, TSC/TRANSP, TOPICS ASTRA. The is simulated GLF23 CDBM05 energy models. injected powers limited negative ion neutral beam, cyclotron electron heating systems. Several plasma...
Time-dependent integrated predictive modelling is carried out using the PTRANSP code to predict fusion power and parameters such as alpha particle density pressure in ITER H-mode plasmas. Auxiliary heating by negative ion neutral beam injection ion-cyclotron of He 3 minority ions are modelled, GLF23 transport model used prediction evolution plasma temperature profiles. Effects steering, torque, rotation, current drive, pedestal temperatures, sawtooth oscillations, magnetic diffusion...
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...