Search

Your search keyword '"Ferron, J. R."' showing total 14 results
Start Over Author "Ferron, J. R." Search Limiters Available in Library Collection
14 results on '"Ferron, J. R."'

1. Analysis of Alfven Eigenmode destabilization in DIII-D high poloidal $\beta$ discharges using a Landau closure model

Author

Varela, J., Spong, D. A., Garcia, L., Huang, J., Murakami, M., Garofalo, A. M., Qian, J. P., Holcomb, C. T., Hyatt, A. W., Ferron, J. R., Collins, C. S., Ren, Q. L., McClenaghan, J., and Guo, W.

Subjects
Physics - Plasma Physics
Abstract

Alfven Eigenmodes are destabilized at the DIII-D pedestal during transient beta drops in high poloidal beta discharges with internal transport barriers (ITBs), driven by n=1 external kink modes, leading to energetic particle losses. There are two different scenarios in the thermal beta recovery phase: with bifurcation (two instability branches with different frequencies) or without bifurcation (single instability branch). We use the reduced MHD equations in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles as well as the geodesic acoustic wave dynamics, to study the properties of the instabilities observed in the DIII-D high poloidal beta discharges and identify the conditions to trigger the bifurcation. The simulations suggest that instabilities with lower frequency in the bifurcation case are ballooning modes driven at the plasma pedestal, while the instability branch with higher frequencies are low n (n<4) Toroidal Alfven Eigenmodes nearby the pedestal. The reverse shear region between the middle and plasma periphery in the non-bifurcated case avoids the excitation of ballooning modes at the pedestal, although Toroidal Alfven Eigenmodes and Reverse Shear Alfven Eigenmodes are unstable in the reverse shear region. The n=1 and n=2 Alfven Eigenmode activity can be suppressed or minimized if the neutral beam injector (NBI) intensity is lower than the experimental value (betaf < 0.03). In addition, if the beam energy or neutral beam injector voltage is lower than in the experiment (Vth,f / VA0 < 0.2), the resonance between beam and thermal plasma is weaker. The n=3,4,5 and 6 AE activity can't be fully suppressed, although the growth rate and frequency is smaller for an optimized neutral beam injector operation regime. In conclusion, AE activity in high poloidal beta discharges can be minimized for optimized NBI operation regimes.

Published
2018
Full Text
View/download PDF

2. Initial off-axis beam experiments in DIII-D

Author

Zeeland, M. A., Heidbrink, W. W., Park, J. M., Prater, R., Holcomb, C. T., Austin, M. E., Ferron, J. R., Greenfield, C. M., Grierson, B. A., Hong, R. -M, Luce, T. C., Mckee, G. R., Moyer, R. A., Murakami, M., Murphy, C. J., Muscatello, C. M., Pace, D. C., Petty, C. C., Rauch, J., Scoville, J. T., Wayne Solomon, and Tobias, B. J.

Published
2012

3. Differences in the H-mode pedestal width of temperature and density

Author

Schneider, P. A., Wolfrum, E., Groebner, R. J., Osborne, T. H., Beurskens, M. N. A., Dunne, M. G., Ferron, J. R., Günter, S., Kurzan, B., Lackner, K., Snyder, P. B., Zohm, H., ASDEX Upgrade Team, DIII-D Team, JET EFDA Contributors, ASDEX Upgrade Team, DIII-D Team, and JET EFDA Contributors

Subjects
Physics, Electron density, Physics::Instrumentation and Detectors, business.industry, Magnetic confinement fusion, Plasma, Condensed Matter Physics, Magnetic field, Computational physics, Optics, Pedestal, Nuclear Energy and Engineering, Physics::Plasma Physics, Electron temperature, business, Scaling, Dimensionless quantity
Abstract

A pedestal database was built using data from type-I ELMy H-modes of ASDEX Upgrade, DIII-D and JET. ELM synchronized pedestal data were analysed with the two-line method. The two-line method is a bilinear fit which shows better reproducibility of pedestal parameters than a modified hyperbolic tangent fit. This was tested with simulated and experimental data. The influence of the equilibrium reconstruction on pedestal parameters was investigated with sophisticated reconstructions from CLISTE and EFIT including edge kinetic profiles. No systematic deviation between the codes could be observed. The flux coordinate system is influenced by machine size, poloidal field and plasma shape. This will change the representation of the width in different coordinates, in particular, the two normalized coordinates ΨN and r/a show a very different dependence on the plasma shape. The scalings derived for the pedestal width, Δ, of all machines suggest a different scaling for the electron temperature and the electron density. Both cases show similar dependence with machine size, poloidal magnetic field and pedestal electron temperature and density. The influence of ion temperature and toroidal magnetic field is different on each of and . In dimensionless form the density pedestal width in ΨN scales with , the temperature pedestal width with . Both widths also show a strong correlation with the plasma shape. The shape dependence originates from the coordinate transformation and is not visible in real space. The presented scalings predict that in ITER the temperature pedestal will be appreciably wider than the density pedestal.

Published
2012

5. Applications of ECH on the DIII-D tokamak and projections for future ECH upgrades.

Author

Prater, R., Buttery, R. J., DeBoo, J., Ferron, J. R., Garofalo, A., Holcomb, C. T., Jackson, G. L., La Haye, R. J., Lohr, J. M., Luce, T. C., Petty, C. C., Politzer, P. A., Solomon, W. M., and Turco, F.

Subjects
ELECTRON cyclotron resonance sources, CYCLOTRONS, POWER electronics, TORQUE, GYROTRONS
Abstract

Electron Cyclotron Heating and Current Drive plays an important role in the DIII-D program. In high performance discharges EC power contributes greatly to MHD stability, and this is particularly important for discharges with low rotational torque applied, as will be the case for ITER. Off-axis EC current drive also plays a key role in the actualization of steady-state scenarios by supporting the desired current profile. In order to carry out these applications at higher beta and higher field, an upgrade of the EC power to 15 MW is needed, and the best gyrotron frequency for the DIII-D program is 117.5 GHz. [ABSTRACT FROM AUTHOR]

Published
2012
Full Text
View/download PDF

6. Physics Operations With the DIII-D Plasma Control System.

Author

Hyatt, A. W., Ferron, J. R., Humphreys, D. A., Chamberlain, F. R., Johnson, R. D., Penaflor, B. G., Piglowski, D. A., Scoville, J. T., and Walker, M. L.

Subjects
*PLASMA confinement, *PHYSICS education, *PHILOSOPHY, *ACTUATORS, *TOKAMAKS
Abstract

The DIII-D device began operation in 1986, and a fully digital plasma control system (PCS) was implemented in 1993. Over time, the success of the PCS to exploit the inherent versatility of the DIII-D device led to a philosophy of using the PCS to control all available plasma system actuators. This made the PCS a very powerful physics tool that is at the core of Physics Operations at DIII-D. The complexity of the DIII-D device and all the systems the PCS must control makes the proper setup of the PCS for new experiments a daunting task. A cadre of physicists specially trained in PCS operation forms the bulk of the Physics Operations staff at DIII-D. They are the interface between experimental plans and successful execution and, as such, are a critical component of each experiment. Physics Operations is also a set of tools and procedures. We will briefly examine some of those tools, such as the TokSys control design and modeling environment and the "smart" PCS setup checklist, that greatly reduce human error in reconfiguring the PCS for a new experiment. We will examine the procedures that allow efficient use of those tools and some of the human factors that can affect productivity. [ABSTRACT FROM AUTHOR]

Published
2010
Full Text
View/download PDF

7. Plasma Startup Design of Fully Superconducting Tokamaks EAST and KSTAR With Implications for ITER.

Author

Leuer, J. A., Eidietis, N. W., Ferron, J. R., Humphreys, D. A., Hyatt, A. W., Jackson, G. L., Johnson, R. D., Penaflor, B. G., Piglowski, D. A., Walker, M. L., Welander, A. S., Yoon, S. W., Hahn, S. H., Oh, Y. K., Xiao, B. J., Wang, H. Z., Yuan, Q. P., and Mueller, D.

Subjects
TOKAMAKS, PLASMA confinement, PLASMA diagnostics, ELECTRIC coils, FUSION reactor walls
Abstract

Recent commissioning of two major fully superconducting (SC)-shaped tokamaks, Experimental Advanced Superconducting Tokamak (EAST) and Korean Superconducting Tokamak Advanced Research (KSTAR), represents a significant advance in magnetic fusion research. The key to commissioning success in these complex and unique tokamaks was as follows: 1) use of a robust, flexible plasma control system (PCS) based on the validated DIII-D design; 2) use of the TokSys design and modeling environment, which is tightly coupled with the DIII-D PCS architecture for first-plasma scenario development and plasma diagnosis; and 3) collaborations with experienced internationally recognized teams of tokamak operations and control experts. We provide an overview of the generic modeling environment and plasma control tools developed and validated within the DIII-D experimental program and applied through an international collaborative program to successfully address the unique constraints associated with the startup of these next-generation tokamaks. The unique characteristics of each tokamak and the machine constraints that must be included in device modeling and simulation, such as SC coil current slew rate limits and the presence of nonlinear magnetic materials, are discussed, along with commissioning and initial operational results. Lessons learned from the startup experience in these devices are summarized, with special emphasis on ramifications for International Thermonuclear Experimental Reactor (ITER). [ABSTRACT FROM AUTHOR]

Published
2010
Full Text
View/download PDF

8. Ramp-Up-Phase Current-Profile Control of Tokamak Plasmas via Nonlinear Programming.

Author

Xu, C., Ou, Y., Dalessio, J., Schuster, E., Luce, T. C., Ferron, J. R., Walker, M. L., and Humphreys, D. A.

Subjects
TOKAMAKS, MAGNETOHYDRODYNAMICS, PARTIAL differential equations, ACTUATORS, HEAT equation
Abstract

The achievement of suitable toroidal-current-density profiles in tokamak plasmas plays an important role in enabling high fusion gain and noninductive sustainment of the plasma current for steady-state operation with improved magnetohydrodynamic stability. The evolution in time of the current profile is related to the evolution of the poloidal magnetic flux, which is modeled in normalized cylindrical coordinates using a partial differential equation (PDE) usually referred to as themagnetic flux diffusion equation. The dynamics of the plasma current density profile can be modified by the total plasma current and the power of the noninductive current drive. These two actuators, which are constrained not only in value and rate but also in their initial and final values, are used to drive the current profile as close as possible to a desired target profile at a specific final time. To solve this constrained finite-time open-loop PDE optimal control problem, model reduction based on proper orthogonal decomposition is combined with sequential quadratic programming in an iterative fashion. The use of a low-dimensional dynamical model dramatically reduces the computational effort and, therefore, the time required to solve the optimization problem, which is critical for a potential implementation of a real-time receding-horizon control strategy. [ABSTRACT FROM AUTHOR]

Published
2010
Full Text
View/download PDF

11. BIBLIOGRAPHY OF SQUID PUBLICATIONS. OCTOBER 1964 - SEPTEMBER 1965.

Author

VIRGINIA UNIV CHARLOTTESVILLE, Alblow, C. M., Fay, J. A., Fenn, J. B., Ferron, J. R., Field, F. H., VIRGINIA UNIV CHARLOTTESVILLE, Alblow, C. M., Fay, J. A., Fenn, J. B., Ferron, J. R., and Field, F. H.

Abstract

Contents: Properties of Materials (Transport, Structural) -- Gases, Liquids, Solids, Physical Processes -- Heat transfer, Mass transfer and mixing, Fluid flow, Mathematical physics; Chemical Processes -- Reaction kinetics, Thermodynamics and thermochemistry; Combustion Phenomena -- Ignition, Flame propagation rate, Flame structure and stability, Flame spectra and radiation, Miscellaneous; Engines and Propulsive Systems; Instrumentation -- Temperature measurement, Pressure measurement.

Published
1965

14. Next-Generation Plasma Control in the DIII-D Tokamak

Author

D.A. Humphreys, Eugenio Schuster, Marco Ariola, Alfredo Pironti, J.A. Leuer, J.R. Ferron, B.G. Penaflor, Robert D. Johnson, D.A. Piglowski, Michael L. Walker, Walker, M. L., Ferron, J. R., Humphreys, D. A., Johnson, R. D., Leuer, J. A., Penaflor, B. G., Piglowski, D. A., Pironti, Alfredo, Ariola, M., and Schuster, E.

Subjects
Physics, Tokamak, DIII-D, Plasma control system, Nuclear engineering, Multivariable calculus, Mechanical Engineering, Energy Engineering and Power Technology, Plasma, Fusion power, Advanced tokamak, Nuclear Energy and Engineering, Civil and Structural Engineering, law.invention, Electronic stability control, law, Control system, General Materials Science, Plasma control
Abstract

OAK A271 NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK. The advanced tokamak (AT) operating mode which is the principal focus of the DIII-D tokamak requires highly integrated and complex plasma control. Simultaneous high performance regulation of the plasma boundary and internal profiles requires multivariable control techniques to account for the highly coupled influences of equilibrium shape, profile, and stability control. This paper describes progress towards the DIII-D At mission goal through both significantly improved real-time computational hardware and control algorithm capability.

Published
2003

Catalog

Books, media, physical & digital resources
See catalog results