Modeling of the Recycling Particle Flux and Electron Particle Transport in the DIII-D Tokamak PDF Download
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Languages : en
Pages : 17
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One of the most difficult aspects of performing an equilibrium particle transport analysis in a diverted tokamak is the determination of the particle flux which enters the plasma after recycling from the divertor plasma, the divertor target plates or the vessel wall. An approach which has been utilized in the past is to model the edge, scrape-off layer (SOL), and divertor plasma to match measured plasma parameters and then use a neutral transport code to obtain an edge recycling flux while trying to match the measured divertor D(x emissivity. Previous simulations were constrained by electron density (n{sub e}) and temperature (T{sub e}), ion temperature (T{sub i}) data at the outer midplane, divertor heat flux from infrared television cameras, and n{sub e}, T{sub e} and particle flux at the target from fixed Langmuir probes, along with the divertor D{sub {alpha}} emissivity. In this paper, we present results of core fueling calculations from the 2-D modeling for ELM-free discharges, constrained by data from the new divertor diagnostics. In addition, we present a simple technique for estimating the recycling flux just after the L-H transition and demonstrate how this technique is supported by the detailed modeling. We will show the effect which inaccuracies in the recycling flux have on the calculated particle flux in the plasma core. For some specific density profiles, it is possible to separate the convective flux from the conductive flux. The diffusion coefficients obtained show a sharp decrease near a normalized radius of 0.9 indicating the presence of a transport barrier.
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Languages : en
Pages : 17
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One of the most difficult aspects of performing an equilibrium particle transport analysis in a diverted tokamak is the determination of the particle flux which enters the plasma after recycling from the divertor plasma, the divertor target plates or the vessel wall. An approach which has been utilized in the past is to model the edge, scrape-off layer (SOL), and divertor plasma to match measured plasma parameters and then use a neutral transport code to obtain an edge recycling flux while trying to match the measured divertor D(x emissivity. Previous simulations were constrained by electron density (n{sub e}) and temperature (T{sub e}), ion temperature (T{sub i}) data at the outer midplane, divertor heat flux from infrared television cameras, and n{sub e}, T{sub e} and particle flux at the target from fixed Langmuir probes, along with the divertor D{sub {alpha}} emissivity. In this paper, we present results of core fueling calculations from the 2-D modeling for ELM-free discharges, constrained by data from the new divertor diagnostics. In addition, we present a simple technique for estimating the recycling flux just after the L-H transition and demonstrate how this technique is supported by the detailed modeling. We will show the effect which inaccuracies in the recycling flux have on the calculated particle flux in the plasma core. For some specific density profiles, it is possible to separate the convective flux from the conductive flux. The diffusion coefficients obtained show a sharp decrease near a normalized radius of 0.9 indicating the presence of a transport barrier.
Author: R. Cohen
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Languages : en
Pages : 16
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The edge-plasma profiles and fluxes to the divertor and walls of a divertor tokamak with a magnetic X-point are simulated by coupling a 2D transport code (UEDGE) and a 3D turbulence code (BOUT). An relaxed iterative coupling scheme is used where each code is run on its characteristic time scale, resulting in a statistical steady state. Plasma variables of density, parallel velocity, and separate ion and electron temperatures are included, together with a fluid neutral model for recycling neutrals at material surfaces. Results for the DIII-D tokamak parameters show that the turbulence is preferentially excited in the outer radial region of the edge where magnetic curvature is destabilizing and that substantial plasma particle flux is transported to the main chamber walls. These results are qualitatively consistent with some experimental observations. The coupled transport/turbulence simulation technique provides a strategy to understanding edge-plasma physics in more detailed than previously available and to significantly enhance the realism of predictions of the performance of future devices.
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Languages : en
Pages : 13
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The edge-plasma profiles and fluxes to the divertor and walls of a divertor tokamak with a magnetic X-point are simulated by coupling a 2D transport code (UEDGE) and a 3D turbulence code (BOUT). An relaxed iterative coupling scheme is used where each code is run on its characteristic time scale, resulting in a statistical steady state. Plasma variables of density, parallel velocity, and separate ion and electron temperatures are included, together with a fluid neutral model for recycling neutrals at material surfaces. Results for the DIII-D tokamak parameters show that the turbulence is preferentially excited in the outer radial region of the edge where magnetic curvature is destabilizing and that substantial plasma particle flux is transported to the main chamber walls. These results are qualitatively consistent with some experimental observations. The coupled transport/turbulence simulation technique provides a strategy to understanding edge-plasma physics in more detailed than previously available and to significantly enhance the realism of predictions of the performance of future devices.
Author: Z. Abou-Assaleh
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Languages : en
Pages : 5
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Languages : en
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We use a model for negative central shear (NCS) heat transport which has a parametric dependence on the plasma conditions with a transport barrier dependence on the minimum of the safety factor profile, 4, qualitatively consistant with experimental observations. Our intention is not to do a detailed investigation of transport models but rather to provide a reasonable model of heat conductivity to be able to simulate effects of electron cyclotron heating (ECH) and current drive (ECCD) on confinement in NCS configurations. We adjust free parameters (c, cl and c2) in the model to obtain a reasonable representation of the temporal evolution of electron and ion temperature profiles consistent with those measured in selected DIII-D shots. In all cases, we use the measured density profiles rather than self- consistently solve for particle sources and particle transport at this time In these results, we employ a simple model for the ECH power deposition by providing an externally supplied heat source for the electrons. The heating deposition location and profile are specified as a function of the toroidal flux coordinate to allow us to independently vary the heating dynamics For the results shown here, we assume a Gaussian profile, typically using a width of[delta][rho]=O.O5 ([delta]r-3 cm in minor radius), with[rho] defined as the square root of toroidal flux All powers are interpreted as that absorbed by the plasma Similarly, the current drive location and profile are specified with the total current (integrated over the assumed profile) modeled as I[sub EccD]=[Gamma]PEcH/neR with[Gamma]=0 005T[sub e], providing current drive efficiency consistent with earlier experiments[6] with a dependence on T[sub e], but fixed Z[sub eff] and trapped particle effects in these simulations Future work will integtate the existing TORCH[7] code into this ECH modeling effort.
Author: Quang Thanh Nguyen
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Languages : en
Pages : 238
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Category : Power resources
Languages : en
Pages : 782
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Author: Junpeng Guo
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Category :
Languages : en
Pages : 98
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Languages : en
Pages : 25
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Particle control of both hydrogen and impurity atoms is important in obtaining reproducible discharges with a low fraction of radiated power in the DIII-D tokamak. The main DIII-D plasma facing components are graphite tiles and Inconel. Hydrogenic species desorbed from graphite during a tokamak discharge can be a major fueling source, especially in unconditioned graphite where these species can saturate the surface regions. In this case the recycling coefficient can exceed unity, leading to an uncontrolled density rise. In addition to removing volatile hydrocarbons and oxygen, DIII-D vessel conditioning efforts have been directed at the reduction of particle fueling from the graphite tiles. Conditioning techniques include: baking to (less-than or equal to) 400°C, low power pulsed discharge cleaning, and glow discharges in deuterium, helium, neon, or argon. Helium glow wall conditioning, is now routinely performed before every tokamak discharge. The effects of these techniques on hydrogen recycling and impurity influxes will be presented. The Inconel walls, while not generally exposed to high heat fluxes, nevertheless represent a source of metal impurities which can lead to impurity accumulation in the discharge and a high fraction of radiated power, particularly in H-mode discharges at higher plasma currents, I{sub p}> 1.5 MA. To reduce metal influx a thin ((approximately)100 nm) low Z film has been applied on all plasma facing surfaces in DIII-D. The application of the boron film, referred to as boronization has the additional benefit over a carbon film of further reducing the oxygen influx. Following the first boronization in DIII-D a regime of very high confinement (VH-mode) was observed, characterized by low ohmic target density, low Z{sub eff}, and low radiated power.
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Languages : en
Pages : 39
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OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, they have made significant progress in developing the building blocks needed for AT operation: (1) they have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved [beta]{sub N}H9 e"10 for 4 [tau]{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m, n) = (3,2) neoclassical tearing mode and then increased [beta]{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 [tau]{sub E}) at the same fusion gain parameter of [beta]{sub N}H9/q952 H"0.4 as ITER but at much higher q95 = 4.2. The authors have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 [tau]{sub E}) with constant density and constant radiated power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; (3) they have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points. They have made detailed investigations of the edge pedestal and SOL: (1) Atomic physics and plasma physics both play significant roles in setting the width of the edge density barrier in H-mode; (2) ELM heat flux conducted to the divertor decreases as density increases; (3) Intermittent, bursty transport contributes to cross field particle transport in the scrape-off layer (SOL) of H-mode and, especially, L-mode plasmas.
