397 research outputs found
Theoretical study of space plasmas Final report, 16 Feb. 1964 - 15 Mar. 1965
Interchange stability of Van Allen belt - Effect of resonant magnetic moment violation on trapped particles - Exact solution of universal instabilit
Theoretical studies of space plasmas Summary report, 3 May 1965 - 1 May 1966
Synchrotron radiation, ionospheric currents, auroral bombardment, and plasma instabilitie
Experimental investigation of the fundamental modes of a collisionless plasma Final report, 10 Mar. 1964 - 31 Oct. 1967
Propagation of electron cyclotron waves and effects of low frequency noise in collisionless plasm
Plasma Edge Kinetic-MHD Modeling in Tokamaks Using Kepler Workflow for Code Coupling, Data Management and Visualization
A new predictive computer simulation tool targeting the development of the H-mode pedestal at the plasma edge in tokamaks and the triggering and dynamics of edge localized modes (ELMs) is presented in this report. This tool brings together, in a coordinated and effective manner, several first-principles physics simulation codes, stability analysis packages, and data processing and visualization tools. A Kepler workflow is used in order to carry out an edge plasma simulation that loosely couples the kinetic code, XGC0, with an ideal MHD linear stability analysis code, ELITE, and an extended MHD initial value code such as M3D or NIMROD. XGC0 includes the neoclassical ion-electron-neutral dynamics needed to simulate pedestal growth near the separatrix. The Kepler workflow processes the XGC0 simulation results into simple images that can be selected and displayed via the Dashboard, a monitoring tool implemented in AJAX allowing the scientist to track computational resources, examine running and archived jobs, and view key physics data, all within a standard Web browser. The XGC0 simulation is monitored for the conditions needed to trigger an ELM crash by periodically assessing the edge plasma pressure and current density profiles using the ELITE code. If an ELM crash is triggered, the Kepler workflow launches the M3D code on a moderate-size Opteron cluster to simulate the nonlinear ELM crash and to compute the relaxation of plasma profiles after the crash. This process is monitored through periodic outputs of plasma fluid quantities that are automatically visualized with AVS/Express and may be displayed on the Dashboard. Finally, the Kepler workflow archives all data outputs and processed images using HPSS, as well as provenance information about the software and hardware used to create the simulation. The complete process of preparing, executing and monitoring a coupled-code simulation of the edge pressure pedestal buildup and the ELM cycle using the Kepler scientific workflow system is described in this paper
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Calculation of Neutral Beam Injection into SSPX
The SSPX spheromak experiment has achieved electron temperatures of 350eV and confinement consistent with closed magnetic surfaces. In addition, there is evidence that the experiment may be up against an operational beta limit for Ohmic heating. To test this barrier, there are firm plans to add two 0.9MW Neutral Beam (NB) sources to the experiment. A question is whether the limit is due to instability. Since the deposited Ohmic power in the core is relatively small the additional power from the beams is sufficient to significantly increase the electron temperature. Here we present results of computations that will support this contention. We have developed a new NB module to calculate the orbits of the injected fast fast-ions. The previous computation made heavy use of tokamak ordering which fails for a tight-aspect-ratio device, where B{sub tor} {approx} B{sub pol}. The model calculates the deposition from the NFREYA package [1]. The neutral from the CX deposition is assumed to be ionized in place, a high-density approximation. The fast ions are then assumed to fill a constant angular momentum orbit. And finally, the fast ions immediately assume the form of a dragged down distribution. Transfer rates are then calculated from this distribution function [2]. The differential times are computed from the orbit times and the particle weights in each flux zone (the sampling bin) are proportional to the time spent in the zone. From this information the flux-surface-averaged profiles are obtained and fed into the appropriate transport equation. This procedure is clearly approximate, but accurate enough to help guide experiments. A major advantage is speed: 5000 particles can be processed in under 4s on our fastest LINUX box. This speed adds flexibility by enabling a ''large'' number of predictive studies. Similar approximations, without the accurate orbit calculation presented here, had some success comparing with experiment and TRANSP [3]. Since our procedure does not have multiple CX and relies on disparate time scales, more detailed understanding requires a ''complete'' NB package such as the NUBEAM [4] module, which follows injected fast ions along with their generations until they enter the main thermal distribution
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Spheromak Energy Transport Studies via Neutral Beam Injection
Results from the SSPX spheromak experiment provide strong motivation to add neutral beam injection (NBI) heating. Such auxiliary heating would significantly advance the capability to study the physics of energy transport and pressure limits for the spheromak. This LDRD project develops the physics basis for using NBI to heat spheromak plasmas in SSPX. The work encompasses three activities: (1) numerical simulation to make quantitative predictions of the effect of adding beams to SSPX, (2) using the SSPX spheromak and theory/modeling to develop potential target plasmas suitable for future application of neutral beam heating, and (3) developing diagnostics to provide the measurements needed for transport calculations. These activities are reported in several publications
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ITER Shape Controller and Transport Simulations
We currently use the CORSICA integrated modeling code for scenario studies for both the DIII-D and ITER experiments. In these simulations, free- or fixed-boundary equilibria are simultaneously converged with thermal evolution determined from transport models providing temperature and current density profiles. Using a combination of fixed boundary evolution followed by free-boundary calculation to determine the separatrix and coil currents. In the free-boundary calculation, we use the state-space controller representation with transport simulations to provide feedback modeling of shape, vertical stability and profile control. In addition to a tightly coupled calculation with simulator and controller imbedded inside CORSICA, we also use a remote procedure call interface to couple the CORSICA non-linear plasma simulations to the controller environments developed within the Mathworks Matlab/Simulink environment. We present transport simulations using full shape and vertical stability control with evolution of the temperature profiles to provide simulations of the ITER controller and plasma response
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Wall-confined high beta spheromak
The spheromak could be extended into the high beta regime by supporting the pressure on flux-conserving walls, allowing the plasma to be in a Taylor state with zero pressure gradient and thus stable to ideal and resistive MHD. The concept yields a potentially attractive, pulsed reactor which would require no external magnets. The flux conserver would be shaped to be stable to the tilt and shift instabilities. We envision a plasma which is ohmically ignited at low beta, with the kinetic pressure growing to beta > 1 by fueling from the edge. The flux conserver would be designed such that the magnetic decay time = the fusion burn time. The thermal capacity of the flux conserver and blanket would exceed the fusion yield per discharge, so that they can be cooled steadily. Ignition is estimated to require minimum technology: 30-100 MJ of pulsed power applied at a 0.5 GW rate generates an estimated bum yield > 1 GJ. The concept thus provides an alternate route to a fusion plasma that is MHD stable at high beta, yielding a reactor that is simple and cheap. The major confinement issue is transport due to grad(T), e.g. driven by high beta modes related to the ITG instability
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Stable bootstrap-current driven equilibria for low aspect ratio tokamaks
Low aspect ratio tokamaks can potentially provide a high ratio of plasma pressure to magnetic pressure {beta} and high plasma current I at a modest size, ultimately leading to a high power density compact fusion power plant. For the concept to be economically feasible, bootstrap current must be a major component of the plasma current. A high value of the Troyon factor {beta}{sub N} and strong shaping are required to allow simultaneous operation at high {beta} and high bootstrap current fraction. Ideal magnetohydrodynamic stability of a range of equilibria at aspect ratio 1.4 is systematically explored by varying the pressure profile and shape. The pressure and current profiles are constrained in such a way as to assure complete bootstrap current alignment. Both {beta}{sub N} and {beta} are defined in terms of the vacuum toroidal field. Equilibria with {beta}{sub N} {ge} 8 and {beta} - 35% to 55% exist which are stable to n = {infinity} ballooning modes, and stable to n = 0, 1,2,3 kink modes with a conducting wall. The dependence of {beta} and {beta}{sub N} with respect to aspect ratio is also considered
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Confinement Studies in High Temperature Spheromak Plasmas
Recent results from the SSPX spheromak experiment demonstrate the potential for obtaining good energy confinement (Te > 350eV and radial electron thermal diffusivity comparable to tokamak L-mode values) in a completely self-organized toroidal plasma. A strong decrease in thermal conductivity with temperature is observed and at the highest temperatures, transport is well below that expected from the Rechester-Rosenbluth model. Addition of a new capacitor bank has produced 60% higher magnetic fields and almost tripled the pulse length to 11ms. For plasmas with T{sub e} > 300eV, it becomes feasible to use modest (1.8MW) neutral beam injection (NBI) heating to significantly change the power balance in the core plasma, making it an effective tool for improving transport analysis. We are now developing detailed designs for adding NBI to SSPX and have developed a new module for the CORSICA transport code to compute the correct fast-ion orbits in SSPX so that we can simulate the effect of adding NBI; initial results predict that such heating can raise the electron temperature and total plasma pressure in the core by a factor of two
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