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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Latest Results from DIII-D and Their Implications for Future Devices

The DIII-D Tokamak has become one of the major sources of physics and technology data for the design of future large machines such as ITER and CIT. In large part this is because of its extremely flexible design and the ability to run tests or add new diagnostic devices with a minimum of down-time and expense. Fundamental plasma physics studies are still the major focus of our experimental program, but increasingly, we are looking at new ways to answer the complex fusion technology questions emerging from the design studies for the next-generation devices. Recent results have demonstrated world record beta plasmas, long quiescent H-mode operation, partial noninductive current drive using both neutral beam and rf power injection, single- and double-null divertor operation, and divertor particle- and heat-load management. The recently installed Advanced Divertor hardware will allow critical transport experiments to be run with independent control of the density. It will also permit the concept of dc helicity injection current drive to be tested. This paper summarizes our most important recent findings and also outlines a few of the many interesting problems that are now under study to address fusion technology questions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Coupling characteristics of the ITER relevant lower hybrid antenna in Tore Supra: experiments and modelling

A new concept of lower hybrid antenna for current drive has been proposed for ITER [Bibet et al, Nuclear Fusion 1995]: the Passive Active Multijunction (PAM) antenna that relies on a periodic combination of active and passive waveguides. An actively cooled PAM antenna at 3.7 GHz has been recently installed on the tokamak Tore Supra. The paper summarizes the comprehensive experimental characterization of the linear coupling properties of the PAM antenna to the Tore Supra plasmas. These experimental results are systematically compared with the linear wave coupling theory via the linear ALOHA code. Good agreement between experimental results and ALOHA have been obtained. The detailed validation of the coupling modelling is an important step toward the validation of the PAM concept in view of further optimizing the electromagnetic properties of the future ITER antenna.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2011

Modelling of Ion Cyclotron Wall Conditioning plasmas

Ion Cyclotron Wall Conditioning (ICWC) is envisioned in ITER to clean the wall from impurities, to control the wall isotopic ratio and the hydrogen recycling in the presence of the toroidal magnetic field. Various experiments and modelling are advancing to consolidate this technique. In this contribution the modeling of ICWC is presented, which can be divided in two parts: plasma description and plasma wall interaction. Firstly a 0D plasma model, based on a set of energy and particle balance equations for Maxwellian Hydrogen and Helium species, is presented. The model takes into account elementary collision processes, coupled RF power, particle confinement, wall recycling, and active gas injection and pumping. The RF plasma production process is based mainly on electron collisional ionization. The dependency of the plasma parameters, the Hydrogen and Helium partial pressures and neutral or ionic fluxes on pressure and RF power are quantitatively in good agreement with those obtained experimentally on TORE SUPRA. Secondly an extension of the 0D model including the description of the wall interaction is presented and compared to TORE SUPRA multi-pulse ICWC discharges.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Behaviors of impurity in ITER plasma with standard type I ELMy H-mode and steady-state scenarios

Self-consistent simulations of impurity behaviors in ITER plasmas in standard Type I ELMy H-mode and steady-state scenarios are investigated using 1.5D BALDUR integrated predictive modeling code. In these simulations, the plasma core transports, including electron and ion thermal, hydrogenic and impurity transports, are predicted using a linear combination of anomalous and neoclassical transports. An anomalous transport is calculated using a theory-based Multimode (MMM95) model; while the neoclassical transport is calculated using NCLASS model. The temperature and density boundary conditions are described at the top of the pedestal. Two different models for hydrogenic and impurity boundary density conditions are considered. The first model is called a 'static boundary density model,' in which the hydrogenic and impurity densities at the boundary are fixed. For the second model, called a 'dynamic boundary density model,' the hydrogenic and impurity densities at the boundary are assumed to be a large fraction of its line-averaged density. For simplicity, the pedestal temperature is assumed to be a constant in all simulations. The combination of a core transport model together with the boundary density models is used to simulate the time evolution of plasma current, temperature, and density profiles for ITER plasmas in standard type I ELMy H-mode and steady-state scenarios. As a result, the behaviors of impurity in ITER plasmas can be investigated. It is found in both ITER scenarios that the total amount of impurity, including beryllium and helium, in plasma core increases rapidly in early state and reaches a steady-state value. The level of impurity content in the steady state depends sensitively on the impurity boundary conditions. The effective charge at the edge is found to be about 1.4 and 1.1 using a static boundary density model and a dynamic boundary density model, respectively. It is also found that the hydrogenic and impurity transports in ITER plasmas for both scenarios is dominated by the kinetic ballooning modes, while the ITG and TEM modes provide the largest contributions for both thermal transports in most of region. In addition, a sensitivity study is carried out to investigate the impacts of pedestal temperature, pedestal density and line-averaged density on the impurity behaviors. It is found that increasing the pedestal temperature results in a reduction of the impurity content. On the other hand, increasing the pedestal density, line-averaged density or impurity influx result in an increase of the impurity content.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Rotational stabilization of resistive wall modes in ITER advanced tokamak scenarios

Rotational stabilization of n=1 resistive wall modes in ITER advanced scenarios [K. Ikeda, Nucl. Fusion 47 (2007)] is investigated, where n is the toroidal mode number. In particular, we present numerical results for the ITER strongly reversed shear case, in comparison to the weakly reversed shear case. The rotation frequency is assumed to be modestly low. Our investigation employs the adaptive eigenfunction independent solution-kinetic (AEGIS-K) code [L. J. Zheng et al., 'AEGIS-K code for linear kinetic analysis of toroidally axisymmetric plasma stability', J. Comput. Phys. (to be published)], which provides a fully kinetic (nonhybrid) and self-consistent (nonperturbative) description. AEGIS-K includes wave-particle resonances, shear Alfven continuum damping, trapped particle effects, and parallel electric effects, but not finite Larmor radius effects. In the case without rotation and kinetic effects included, we find that the strongly reversed shear configuration is more favorable for perfectly conducting wall stabilization of resistive wall modes, in that it has a higher conducting wall beta limit than the weakly reversed shear case. With sufficient rotation, the strongly reversed shear case can also achieve a higher beta limit for completely suppressing the resistive wall modes. However, the marginal rotation frequency required for complete resistive wall mode stabilization in the strongly reversed shear case is about twice as high as that required for the weakly reversed shear case.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Understanding and predicting the dynamics of tokamak discharges during startup and rampdown

Understanding the dynamics of plasma startup and termination is important for present tokamaks and for predictive modeling of future burning plasma devices such as ITER. We report on experiments in the DIII-D tokamak that explore the plasma startup and rampdown phases and on the benchmarking of transport models. Key issues have been examined such as plasma initiation and burnthrough with limited inductive voltage and achieving flattop and maximum burn within the technical limits of coil systems and their actuators while maintaining the desired q profile. Successful rampdown requires scenarios consistent with technical limits, including controlled H-L transitions, while avoiding vertical instabilities, additional Ohmic transformer flux consumption, and density limit disruptions. Discharges were typically initiated with an inductive electric field typical of ITER, 0.3 V/m, most with second harmonic electron cyclotron assist. A fast framing camera was used during breakdown and burnthrough of low Z impurity charge states to study the formation physics. An improved 'large aperture' ITER startup scenario was developed, and aperture reduction in rampdown was found to be essential to avoid instabilities. Current evolution using neoclassical conductivity in the CORSICA code agrees with rampup experiments, but the prediction of the temperature and internal inductance evolution using the Coppi-Tang model for electron energy transport is not yet accurate enough to allow extrapolation to future devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Plasma wave simulation based on a versatile finite element method solver

The possibility of plasma wave simulation based on the finite element method (FEM) was explored, and a new full wave simulation code of the lower hybrid (LH) wave, Lower Hybrid wavE Analysis based on FEM (LHEAF), was developed. The efficient iterative solver for the LH wave [O. Meneghini, S. Shiraiwa, and R. Parker, Phys. Plasmas 16, 090701 (2009)] was coupled to one-dimensional Fokker-Planck calculation to obtain the self-consistent distribution function. In LHEAF, seamless handling of the core, scrape off layer, and antenna regions were realized and the wave launching structure were naturally included into the simulation. Simulation results of a Maxwellian tokamak plasma showed good agreement with ray tracing calculations and the TORIC-LH spectral solver [J. C. Wright, P. T. Bonoli, A. E. Schmidt et al., Phys. Plasmas 16, 072502 (2009)]. Compared to spectral domain solvers, the computational requirements are reduced significantly, allowing an Alcator C [M. Porkolab, J. J. Schuss, B. Lloyd et al., Phys. Rev. Lett. 53, 450 (1984)] scale plasma simulation on a desktop computer. Also, the LH full wave simulation of an ITER scale plasma was demonstrated for the first time with moderate increase in the problem size. In addition, the flexibility of the FEM approach has been exploited to address issues of antenna-plasma coupling in the LH and ion cyclotron range of frequencies (ICRF). Techniques using the FEM package for this purpose were validated on the traditional grill antenna and have been applied to the interdigital-line antenna of LH wave. Application to the near field analysis of the new Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda et al., Phys. Plasmas 1, 1511 (1994)] ICRF antenna is in progress.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

To the theory of high-power gyrotrons with uptapered resonators

In high-power gyrotrons it is desirable to combine an optimal resonator length with the optimal value of the resonator quality factor. In resonators with the constant radius of the central part, the possibilities of this combination are limited because the quality factor of the resonator sharply increases with its length. Therefore the attempts to increase the length for maximizing the efficiency leads to such increase in the quality factor which makes the optimal current too small. Resonators with slightly uptapered profiles offer more flexibility in this regard. In such resonators, one can separate optimization of the interaction length from optimization of the quality factor because the quality factor determined by diffractive losses can be reduced by increasing the angle of uptapering. In the present paper, these issues are analyzed by studying as a typical high-power 17 GHz gyrotron which is currently under development in Europe for ITER (http://en.wikipedia.org/wiki/ITER). The effect of a slight uptapering of the resonator wall on the efficiency enhancement and the purity of the radiation spectrum in the process of the gyrotron start-up and power modulation are studied. Results show that optimal modification of the shape of a slightly uptapered resonator may result in increasing the gyrotron power from 1052 to 1360 kW.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Erosion, Contamination, and Migration

This paper will summarize studies of carbon impurity sources, contamination, and migration developed through JET methane gas injection experiments. These studies were analyzed using the 2D SOL code EDGE2D/NIMBUS. The code is capable of repeating the JET analysis using the ITER geometry and SOL plasma. This allows assessment of whether the physical processes occurring in JET might also occur in ITER, and thus whether the JET results transfer, in any sense, to the ITER plasmas. Certainly, the ITER choice of wall materials (W and Be) is different than for the present JET C studies. So the present status of these studies is to relate JET carbon behavior to carbon in ITER.JET carbon sources were studied spectroscopically and analyzed with atomic physics models in EDGE2D. The carbon sources are dominated by chemical sputtering at rates which are within a factor-of-two of the published literature. The JET carbon contamination is dominated by main chamber sources which are ionized in the main chamber SOL about 1-2 cm from the separatrix. Contamination occurs from carbon ions which diffuse across the field lines and reach the separatrix before they can parallel transport to the divertor. JET carbon migration was studied by injecting methane composed of {sup 13}C on the last run day before an opening and then analyzing removed tiles to identify migration to those locations. Modeling was accomplished by the same EDGE2D models that were used to describe the carbon sources and contamination. The entire migration process is complicated.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Modeling of Erosion and Deposition on Plasma Facing Walls

The unavoidable contact of plasmas with surrounding walls results in plasma-surface interactions (PSIs) that are strongly interlinked and cannot be studied separately. Computer modeling has become increasingly important in understanding mechanisms of PSIs in present devices, ITER and beyond. Modeling of erosion and deposition requires self-consistent calculations of (1) erosion of the wall surface, (2) transport of eroded impurities in the plasma above the surface, (3) redeposition of returning impurities on the surface and (4) resultant material mixing below the surface. In addition, it is necessary to use exact rate coefficients for collision reactions in the plasma and related data for the surface reactions on plasma-facing walls. This chapter describes modeling codes in terms of such PSI issues and the physical and chemical bases of the interactions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Benefits and Challenges of the Use of High-Z Plasma Facing Materials in Fusion Devices

The use of high-Z plasma facing components requires intensive research in all areas, i.e. in plasma wall-interaction, in the physics of the confined plasma, diagnostic, and in material development. Only a few present day divertor tokamaks - mainly Alcator C-Mod and ASDEX Upgrade - gained experience with the refractory metals molybdenum and tungsten, respectively. ASDEX Upgrade was stepwise converted from graphite to tungsten PFCs and in parallel a reduction of the deuterium retention by almost a factor of ten has been observed due to the strong suppression of D co-deposition with carbon. The deuterium retained in W is in line with laboratory results. In order to diagnose W sources and the W content in the main plasma adequate spectroscopic methods had to be developed. As expected from the sputtering threshold of Mo and W, negligible erosion by the thermal divertor background plasma is found in ASDEX Upgrade and Alcator C-Mod under low temperature divertor conditions. However, erosion by fast particles and intrinsic impurities, which additionally might be accelerated in rectified electrical fields observed during ion cyclotron frequency heating, plays an important role. The Mo and W concentrations in the plasma centre are strongly affected by plasma transport and variations up to a factor of 50 are observed for similar influxes. However, it could be demonstrated that sawteeth and turbulent transport driven by central heating can suppress central accumulation. The inward transport of high-Z ions at the edge can be efficiently reduced by 'flushing' the pedestal region caused by frequent edge instabilities. Since with metal walls the edge radiation by low-Z impurities is reduced, it has to be substituted in a pure high-Z device by artificially injected low-Z impurities in order to keep the power load at an acceptable level. Experiments at ASDEX Upgrade suggest that a regime with benign erosion and favourable confinement can be achieved. Extrapolations to ITER and DEMO are difficult since the physics of plasma transport is not yet completely understood, the particle and energy fluxes are orders of magnitude higher and the technical boundary conditions in DEMO strongly differ from those of present day devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Mixed Material Plasma-Surface Interactions in ITER: Recent Results from the PISCES Group

This paper summarizes recent PISCES studies focused on the effects associated with mixed species plasmas that are similar in composition to what one might expect in ITER. Formation of nanometer scale whiskerlike features occurs in W surfaces exposed to pure He and mixed D/He plasmas and appears to be associated with the formation of He nanometer-scaled bubbles in the W surface. Studies of Be-W alloy formation in Be-seeded D plasmas suggest that this process may be important in ITER all metal wall operational scenarios. Studies also suggest that BeD formation via chemical sputtering of Be walls may be an important first wall erosion mechanism. D retention in ITER mixed materials has also been studied. The D release behavior from beryllium co-deposits does not appear to be a diffusion dominated process, but instead is consistent with thermal release from a number of variable trapping energy sites. As a result, the amount of tritium remaining in codeposits in ITER after baking will be determined by the maximum temperature achieved, rather than by the duration of the baking cycle.

36 MATERIALS SCIENCE↗
DOE OSTI2010

Material Mixing of Tungsten with Carbon and Helium

In ITER, graphite and tungsten are used for divertor materials and are mixed through erosion, transport, and redeposition. Helium, a fusion reactant, is an intrinsic element in fusion plasmas that impinges on the metallic wall materials to form He bubbles. W-C mixed layers and He bubble layers greatly affect tritium retention. In this paper, impacts of W-C material mixing on erosion and hydrogen isotope retention are reviewed. Then, recent results on carbon deposition on tungsten in TEXTOR tokamak and helium effects on blistering and retention are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Wall forces produced during ITER disruptions

Nonlinear simulations with the M3D code [W. Park et al., Phys. Plasmas 6, 1796 (1999)] are performed of disruptions produced by large scale magnetohydrodynamic instabilities. The toroidally asymmetric wall forces produced during a disruption are calculated in an ITER [T. Hender et al., Nucl. Fusion 47, S128 (2007)] model. The disruption is produced by a vertical displacement event and a kink mode. Expressions are derived for the wall force, including the sideways force, using a thin conducting wall model. The scaling of wall force with {gamma}{tau}{sub w} is obtained, where {gamma} is the kink growth rate and {tau}{sub w} is the wall penetration time. The largest force occurs with {gamma}{tau}{sub w{approx_equal}}1. A theory is developed of the wall force produced by kink modes. The theory is in qualitative agreement with the simulations and Joint European Torus [V. Riccardo et al., Nucl. Fusion 49, 055012 (2009)] experiments. In particular, the theory and simulations give dependence of the sideways on {gamma}{tau}{sub w}, correlation of sideways force with sideways plasma displacement, and correlation of toroidally varying plasma current with toroidally varying vertical displacement.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

The role of kinetic effects, including plasma rotation and energetic particles, in resistive wall mode stability

The resistive wall mode (RWM) instability in high-beta tokamaks is stabilized by energy dissipation mechanisms that depend on plasma rotation and kinetic effects. Kinetic modification of ideal stability calculated with the 'MISK' code [B. Hu et al., Phys. Plasmas 12, 057301 (2005)] is outlined. For an advanced scenario ITER [R. Aymar et al., Nucl. Fusion 41, 1301 (2001)] plasma, the present calculation finds that alpha particles are required for RWM stability at presently expected levels of plasma rotation. Kinetic stabilization theory is tested in an experiment in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)] that produced marginally stable plasmas with various energetic particle contents. Plasmas with the highest and lowest energetic particle content agree with calculations predicting that increased energetic particle pressure is stabilizing but does not alter the nonmonotonic dependence of stability on plasma rotation due to thermal particle resonances. Presently, the full MISK model, including thermal particles and an isotropic slowing-down distribution function for energetic particles, overpredicts stability in NSTX experiments. Minor alteration of either effect in the theory may yield agreement; several possibilities are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Linear gyrokinetic simulation of high-n toroidal Alfven eigenmodes in a burning plasma

A hybrid gyrokinetic ions/massless fluid electron model is used to study the stability of high-n toroidal Alfven eigenmodes (TAEs) in ITER [M. Shimada et al., Nucl. Fusion 47, S1 (2007)]. The hybrid model has been implemented in the particle-in-cell turbulence simulation code GEM [Y. Chen and S. E. Parker, J. Comput. Phys. 220, 839 (2007)]. The adequacy of the hybrid model for simulating TAEs has been previously demonstrated [J. Lang et al., Phys. Plasmas 16, 102101 (2009)] by comparing the simulated TAE mode frequency and structure with an eigenmode analysis, and the thermal ion kinetic damping effect with analytic theory. By using a global particle-in-cell code the effects of large orbit width and nonlocal mode structures can be accurately included. Damping rate due to numerical filtering is carefully monitored, and convergence with respect to particle number, grid resolution, etc., is thoroughly tested. The simulations show that the most unstable modes in ITER lie in the rage of 10<n<20. Thermal ion pressure effect and alpha particle nonperturbative effect are important in determining the mode radial location and stability threshold. The thermal ion Landau damping rate and radiative damping rate from the simulations are compared with analytical estimates. The thermal ion Landau damping is the dominant damping mechanism. Plasma elongation has a strong stabilizing effect on the alpha driven TAEs. The central alpha particle pressure threshold for the most unstable n=15 mode is about {beta}{sub {alpha}(0)}=0.7% for the fully shaped ITER equilibrium.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Modeling of Hydrogen Line Shapes for the Diagnostic of ITER

The state of art of the line shape modeling techniques involved in tokamak edge plasma spectroscopy is reported, in the context of the preparation for ITER. Hydrogen spectra are calculated assuming a line-of-sight crossing a 2D-plasma background obtained from numerical simulations. The Doppler, Zeeman and Stark effects are retained. Ion dynamics effects are accounted for by using the numerical simulation method. The possibility for a line shape-based diagnostic of the ITER divertor plasma is examined through fittings of simulated spectra and comparison with the input plasma fields.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗