Gyrokinetic particle simulations of the effects of compressional magnetic perturbations on drift-Alfvenic instabilities in tokamaks

DOE PAGES

Dong, Ge; Bao, Jian; Bhattacharjee, Amitava; ...

2017-08-10

The compressional component of magnetic perturbation δB- || to can play an important role in drift-Alfvenic instabilities in tokamaks, especially as the plasma β increases (β is the ratio of kinetic pressure to magnetic pressure). In this work, we have formulated a gyrokinetic particle simulation model incorporating δB- ||, and verified the model in kinetic Alfven wave simulations using the Gyrokinetic Toroidal Code in slab geometry. Simulations of drift-Alfvenic instabilities in tokamak geometry shows that the kinetic ballooning mode (KBM) growth rate decreases more than 20% when δB- || is neglected for β e = 0.02, and that δB- ||more » to has stabilizing effects on the ion temperature gradient instability, but negligible effects on the collisionless trapped electron mode. Lastly, the KBM growth rate decreases about 15% when equilibrium current is neglected.« less

Full-f version of GENE for turbulence in open-field-line systems

NASA Astrophysics Data System (ADS)

Pan, Q.; Told, D.; Shi, E. L.; Hammett, G. W.; Jenko, F.

2018-06-01

Unique properties of plasmas in the tokamak edge, such as large amplitude fluctuations and plasma-wall interactions in the open-field-line regions, require major modifications of existing gyrokinetic codes originally designed for simulating core turbulence. To this end, the global version of the 3D2V gyrokinetic code GENE, so far employing a δf-splitting technique, is extended to simulate electrostatic turbulence in straight open-field-line systems. The major extensions are the inclusion of the velocity-space nonlinearity, the development of a conducting-sheath boundary, and the implementation of the Lenard-Bernstein collision operator. With these developments, the code can be run as a full-f code and can handle particle loss to and reflection from the wall. The extended code is applied to modeling turbulence in the Large Plasma Device (LAPD), with a reduced mass ratio and a much lower collisionality. Similar to turbulence in a tokamak scrape-off layer, LAPD turbulence involves collisions, parallel streaming, cross-field turbulent transport with steep profiles, and particle loss at the parallel boundary.

Runaway Electrons Modeling and Nanoparticle Plasma Jet Penetration into Tokamak Plasma

NASA Astrophysics Data System (ADS)

Galkin, S. A.; Bogatu, I. N.

2017-10-01

A novel idea to probe runaway electrons (REs) by superfast injection of high velocity nanoparticle plasma jet (NPPJ) from a plasma accelerator needs to be sustained by both RE dynamics modeling and simulation of NPPJ penetration through increasing tokamak magnetic field. We present our recent progress in both areas. RE simulation is based on the model, including Dreicer and ``avalanche'' mechanisms of RE generation, with emphasis on high Zeff effects. The high-density hyper-velocity C60 and BN NPPJ penetration through transversal B-field is conducted with the Hybrid Electro-Magnetic code (HEM-2D) in cylindrical coordinates, with 1/R B-field dependence for both DIII-D and ITER tokamaks. Work is supported in part by US DOE SBIR Grant.

Global linear gyrokinetic particle-in-cell simulations including electromagnetic effects in shaped plasmas

NASA Astrophysics Data System (ADS)

Mishchenko, A.; Borchardt, M.; Cole, M.; Hatzky, R.; Fehér, T.; Kleiber, R.; Könies, A.; Zocco, A.

2015-05-01

We give an overview of recent developments in electromagnetic simulations based on the gyrokinetic particle-in-cell codes GYGLES and EUTERPE. We present the gyrokinetic electromagnetic models implemented in the codes and discuss further improvements of the numerical algorithm, in particular the so-called pullback mitigation of the cancellation problem. The improved algorithm is employed to simulate linear electromagnetic instabilities in shaped tokamak and stellarator plasmas, which was previously impossible for the parameters considered.

Simulations of Tokamak Edge Turbulence Including Self-Consistent Zonal Flows

NASA Astrophysics Data System (ADS)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge is summarized in this mini-conference talk. A more detailed report on this work is presented in a poster at this conference. This work extends our previous work to include self-consistent zonal flows and their effects. The previous work addressed the simulation of L-mode tokamak edge turbulence using the turbulence code BOUT. The calculations used realistic single-null geometry and plasma parameters of the DIII-D tokamak and produced fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the US Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Kinetic simulation of edge instability in fusion plasmas

NASA Astrophysics Data System (ADS)

Fulton, Daniel Patrick

In this work, gyrokinetic simulations in edge plasmas of both tokamaks and field reversed. configurations (FRC) have been carried out using the Gyrokinetic Toroidal Code (GTC) and A New Code (ANC) has been formulated for cross-separatrix FRC simulation. In the tokamak edge, turbulent transport in the pedestal of an H-mode DIII-D plasma is. studied via simulations of electrostatic driftwaves. Annulus geometry is used and simulations focus on two radial locations corresponding to the pedestal top with mild pressure gradient and steep pressure gradient. A reactive trapped electron instability with typical ballooning mode structure is excited in the pedestal top. At the steep gradient, the electrostatic instability exhibits unusual mode structure, peaking at poloidal angles theta=+- pi/2. Simulations find this unusual mode structure is due to steep pressure gradients in the pedestal but not due to the particular DIII-D magnetic geometry. Realistic DIII-D geometry has a stabilizing effect compared to a simple circular tokamak geometry. Driftwave instability in FRC is studied for the first time using gyrokinetic simulation. GTC. is upgraded to treat realistic equilibrium calculated by an MHD equilibrium code. Electrostatic local simulations in outer closed flux surfaces find ion-scale modes are stable due to the large ion gyroradius and that electron drift-interchange modes are excited by electron temperature gradient and bad magnetic curvature. In the scrape-off layer (SOL) ion-scale modes are excited by density gradient and bad curvature. Collisions have weak effects on instabilities both in the core and SOL. Simulation results are consistent with density fluctuation measurements in the C-2 experiment using Doppler backscattering (DBS). The critical density gradients measured by the DBS qualitatively agree with the linear instability threshold calculated by GTC simulations. One outstanding critical issue in the FRC is the interplay between turbulence in the FRC. core and SOL regions. While the magnetic flux coordinates used by GTC provide a number of computational advantages, they present unique challenges at the magnetic field separatrix. To address this limitation, a new code, capable of coupled core-SOL simulations, is formulated, implemented, and successfully verified.

Study of the L-mode tokamak plasma “shortfall” with local and global nonlinear gyrokinetic δf particle-in-cell simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chowdhury, J.; Wan, Weigang; Chen, Yang

2014-11-15

The δ f particle-in-cell code GEM is used to study the transport “shortfall” problem of gyrokinetic simulations. In local simulations, the GEM results confirm the previously reported simulation results of DIII-D [Holland et al., Phys. Plasmas 16, 052301 (2009)] and Alcator C-Mod [Howard et al., Nucl. Fusion 53, 123011 (2013)] tokamaks with the continuum code GYRO. Namely, for DIII-D the simulations closely predict the ion heat flux at the core, while substantially underpredict transport towards the edge; while for Alcator C-Mod, the simulations show agreement with the experimental values of ion heat flux, at least within the range of experimental error.more » Global simulations are carried out for DIII-D L-mode plasmas to study the effect of edge turbulence on the outer core ion heat transport. The edge turbulence enhances the outer core ion heat transport through turbulence spreading. However, this edge turbulence spreading effect is not enough to explain the transport underprediction.« less

Numerical Simulation of MIG for 42 GHz, 200 kW Gyrotron

NASA Astrophysics Data System (ADS)

Singh, Udaybir; Bera, Anirban; Kumar, Narendra; Purohit, L. P.; Sinha, Ashok K.

2010-06-01

A triode type magnetron injection gun (MIG) of a 42 GHz, 200 kW gyrotron for an Indian TOKAMAK system is designed by using the commercially available code EGUN. The operating voltages of the modulating anode and the accelerating anode are 29 kV and 65 kV respectively. The operating mode of the gyrotron is TE03 and it is operated in fundamental harmonic. The simulated results of MIG obtained with the EGUN code are validated with another trajectory code TRAK.

Kinetic simulations of scrape-off layer physics in the DIII-D tokamak

DOE PAGES

Churchill, Randy M.; Canik, John M.; Chang, C. S.; ...

2016-12-27

Simulations using the fully kinetic code XGCa were undertaken to explore the impact of kinetic effects on scrape-off layer (SOL) physics in DIII-D H-mode plasmas. XGCa is a total- f, gyrokinetic code which self-consistently calculates the axisymmetric electrostatic potential and plasma dynamics, and includes modules for Monte Carlo neutral transport. Fluid simulations are normally used to simulate the SOL, due to its high collisionality. However, depending on plasma conditions, a number of discrepancies have been observed between experiment and leading SOL fluid codes (e.g. SOLPS), including underestimating outer target temperatures, radial electric field in the SOL, parallel ion SOL flowsmore » at the low field side, and impurity radiation. Many of these discrepancies may be linked to the fluid treatment, and might be resolved by including kinetic effects in SOL simulations. The XGCa simulation of the DIII-D tokamak in a nominally sheath-limited regime show many noteworthy features in the SOL. The density and ion temperature are higher at the low-field side, indicative of ion orbit loss. The SOL ion Mach flows are at experimentally relevant levels ( Mi ~0.5), with similar shapes and poloidal variation as observed in various tokamaks. Surprisingly, the ion Mach flows close to the sheath edge remain subsonic, in contrast to the typical fluid Bohm criterion requiring ion flows to be above sonic at the sheath edge. Related to this are the presence of elevated sheath potentials, eΔΦ/T e ~ 3–4, over most of the SOL, with regions in the near-SOL close to the separatrix having eΔΦ/Te > 4. Finally, these two results at the sheath edge are a consequence of non-Maxwellian features in the ions and electrons there.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dey, Ritu; Ghosh, Joydeep; Chowdhuri, M. B.

Neutral particle behavior in Aditya tokamak, which has a circular poloidal ring limiter at one particular toroidal location, has been investigated using DEGAS2 code. The code is based on the calculation using Monte Carlo algorithms and is mainly used in tokamaks with divertor configuration. This code has been successfully implemented in Aditya tokamak with limiter configuration. The penetration of neutral hydrogen atom is studied with various atomic and molecular contributions and it is found that the maximum contribution comes from the dissociation processes. For the same, H α spectrum is also simulated which was matched with the experimental one. Themore » dominant contribution around 64% comes from molecular dissociation processes and neutral particle is generated by those processes have energy of ~ 2.0 eV. Furthermore, the variation of neutral hydrogen density and H α emissivity profile are analysed for various edge temperature profiles and found that there is not much changes in H α emission at the plasma edge with the variation of edge temperature (7 to 40 eV).« less

Simulation of tokamak armour erosion and plasma contamination at intense transient heat fluxes in ITER

NASA Astrophysics Data System (ADS)

Landman, I. S.; Bazylev, B. N.; Garkusha, I. E.; Loarte, A.; Pestchanyi, S. E.; Safronov, V. M.

2005-03-01

For ITER, the potential material damage of plasma facing tungsten-, CFC-, or beryllium components during transient processes such as ELMs or mitigated disruptions are simulated numerically using the MHD code FOREV-2D and the melt motion code MEMOS-1.5D for a heat deposition in the range of 0.5-3 MJ/m 2 on the time scale of 0.1-1 ms. Such loads can cause significant evaporation at the target surface and a contamination of the SOL by the ions of evaporated material. Results are presented on carbon plasma dynamics in toroidal geometry and on radiation fluxes from the SOL carbon ions obtained with FOREV-2D. The validation of MEMOS-1.5D against the plasma gun tokamak simulators MK-200UG and QSPA-Kh50, based on the tungsten melting threshold, is described. Simulations with MEMOS-1.5D for a beryllium first wall that provide important details about the melt motion dynamics and typical features of the damage are reported.

Largescale Long-term particle Simulations of Runaway electrons in Tokamaks

NASA Astrophysics Data System (ADS)

Liu, Jian; Qin, Hong; Wang, Yulei

2016-10-01

To understand runaway dynamical behavior is crucial to assess the safety of tokamaks. Though many important analytical and numerical results have been achieved, the overall dynamic behaviors of runaway electrons in a realistic tokamak configuration is still rather vague. In this work, the secular full-orbit simulations of runaway electrons are carried out based on a relativistic volume-preserving algorithm. Detailed phase-space behaviors of runaway electrons are investigated in different timescales spanning 11 orders. A detailed analysis of the collisionless neoclassical scattering is provided when considering the coupling between the rotation of momentum vector and the background field. In large timescale, the initial condition of runaway electrons in phase space globally influences the runaway distribution. It is discovered that parameters and field configuration of tokamaks can modify the runaway electron dynamics significantly. Simulations on 10 million cores of supercomputer using the APT code have been completed. A resolution of 107 in phase space is used, and simulations are performed for 1011 time steps. Largescale simulations show that in a realistic fusion reactor, the concern of runaway electrons is not as serious as previously thought. This research was supported by National Magnetic Connement Fusion Energy Research Project (2015GB111003, 2014GB124005), the National Natural Science Foundation of China (NSFC-11575185, 11575186) and the GeoAlgorithmic Plasma Simulator (GAPS) Project.

Numerical modelling of geodesic acoustic mode relaxation in a tokamak edge

DOE PAGES

Dorf, M. A.; Cohen, R. H.; Dorr, M.; ...

2013-05-08

Here, the edge of a tokamak in a high confinement (H mode) regime is characterized by steep density gradients and a large radial electric field. Recent analytical studies demonstrated that the presence of a strong radial electric field consistent with a subsonic pedestal equilibrium modifies the conventional results of the neoclassical formalism developed for the core region. In the present work we make use of the recently developed gyrokinetic code COGENT to numerically investigate neoclassical transport in a tokamak edge including the effects of a strong radial electric field. The results of numerical simulations are found to be in goodmore » qualitative agreement with the theoretical predictions and the quantitative discrepancy is discussed. In addition, the present work investigates the effects of a strong radial electric field on the relaxation of geodesic acoustic modes (GAMs) in a tokamak edge. Numerical simulations demonstrate that the presence of a strong radial electric field characteristic of a tokamak pedestal can enhance the GAM decay rate, and heuristic arguments elucidating this finding are provided.« less

Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST.

PubMed

Xu, X Q

2008-07-01

We present gyrokinetic neoclassical simulations of tokamak plasmas with a self-consistent electric field using a fully nonlinear (full- f ) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five-dimensional computational grid in phase space. The present implementation is a method of lines approach where the phase-space derivatives are discretized with finite differences, and implicit backward differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving the gyrokinetic Poisson equation with self-consistent poloidal variation. With a four-dimensional (psi,theta,micro) version of the TEMPEST code, we compute the radial particle and heat fluxes, the geodesic-acoustic mode, and the development of the neoclassical electric field, which we compare with neoclassical theory using a Lorentz collision model. The present work provides a numerical scheme for self-consistently studying important dynamical aspects of neoclassical transport and electric field in toroidal magnetic fusion devices.

Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST

NASA Astrophysics Data System (ADS)

Xu, X. Q.

2008-07-01

We present gyrokinetic neoclassical simulations of tokamak plasmas with a self-consistent electric field using a fully nonlinear (full- f ) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five-dimensional computational grid in phase space. The present implementation is a method of lines approach where the phase-space derivatives are discretized with finite differences, and implicit backward differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving the gyrokinetic Poisson equation with self-consistent poloidal variation. With a four-dimensional (ψ,θ,γ,μ) version of the TEMPEST code, we compute the radial particle and heat fluxes, the geodesic-acoustic mode, and the development of the neoclassical electric field, which we compare with neoclassical theory using a Lorentz collision model. The present work provides a numerical scheme for self-consistently studying important dynamical aspects of neoclassical transport and electric field in toroidal magnetic fusion devices.

Progress on the DPASS project

NASA Astrophysics Data System (ADS)

Galkin, Sergei A.; Bogatu, I. N.; Svidzinski, V. A.

2015-11-01

A novel project to develop Disruption Prediction And Simulation Suite (DPASS) of comprehensive computational tools to predict, model, and analyze disruption events in tokamaks has been recently started at FAR-TECH Inc. DPASS will eventually address the following aspects of the disruption problem: MHD, plasma edge dynamics, plasma-wall interaction, generation and losses of runaway electrons. DPASS uses the 3-D Disruption Simulation Code (DSC-3D) as a core tool and will have a modular structure. DSC is a one fluid non-linear, time-dependent 3D MHD code to simulate dynamics of tokamak plasma surrounded by pure vacuum B-field in the real geometry of a conducting tokamak vessel. DSC utilizes the adaptive meshless technique with adaptation to the moving plasma boundary, with accurate magnetic flux conservation and resolution of the plasma surface current. DSC has also an option to neglect the plasma inertia to eliminate fast magnetosonic scale. This option can be turned on/off as needed. During Phase I of the project, two modules will be developed: the computational module for modeling the massive gas injection and main plasma respond; and the module for nanoparticle plasma jet injection as an innovative disruption mitigation scheme. We will report on this development progress. Work is supported by the US DOE SBIR grant # DE-SC0013727.

Tokamak magneto-hydrodynamics and reference magnetic coordinates for simulations of plasma disruptions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zakharov, Leonid E.; Li, Xujing

This paper formulates the Tokamak Magneto-Hydrodynamics (TMHD), initially outlined by X. Li and L. E. Zakharov [Plasma Science and Technology 17(2), 97–104 (2015)] for proper simulations of macroscopic plasma dynamics. The simplest set of magneto-hydrodynamics equations, sufficient for disruption modeling and extendable to more refined physics, is explained in detail. First, the TMHD introduces to 3-D simulations the Reference Magnetic Coordinates (RMC), which are aligned with the magnetic field in the best possible way. The numerical implementation of RMC is adaptive grids. Being consistent with the high anisotropy of the tokamak plasma, RMC allow simulations at realistic, very high plasmamore » electric conductivity. Second, the TMHD splits the equation of motion into an equilibrium equation and the plasma advancing equation. This resolves the 4 decade old problem of Courant limitations of the time step in existing, plasma inertia driven numerical codes. The splitting allows disruption simulations on a relatively slow time scale in comparison with the fast time of ideal MHD instabilities. A new, efficient numerical scheme is proposed for TMHD.« less

Partnership for Edge Physics (EPSI), University of Texas Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moser, Robert; Carey, Varis; Michoski, Craig

Simulations of tokamak plasmas require a number of inputs whose values are uncertain. The effects of these input uncertainties on the reliability of model predictions is of great importance when validating predictions by comparison to experimental observations, and when using the predictions for design and operation of devices. However, high fidelity simulation of tokamak plasmas, particular those aimed at characterization of the edge plasma physics, are computationally expensive, so lower cost surrogates are required to enable practical uncertainty estimates. Two surrogate modeling techniques have been explored in the context of tokamak plasma simulations using the XGC family of plasma simulationmore » codes. The first is a response surface surrogate, and the second is an augmented surrogate relying on scenario extrapolation. In addition, to reduce the costs of the XGC simulations, a particle resampling algorithm was developed, which allows marker particle distributions to be adjusted to maintain optimal importance sampling. This means that the total number of particles in and therefore the cost of a simulation can be reduced while maintaining the same accuracy.« less

Residual zonal flows in tokamaks and stellarators at arbitrary wavelengths

NASA Astrophysics Data System (ADS)

Monreal, Pedro; Calvo, Iván; Sánchez, Edilberto; Parra, Félix I.; Bustos, Andrés; Könies, Axel; Kleiber, Ralf; Görler, Tobias

2016-04-01

In the linear collisionless limit, a zonal potential perturbation in a toroidal plasma relaxes, in general, to a non-zero residual value. Expressions for the residual value in tokamak and stellarator geometries, and for arbitrary wavelengths, are derived. These expressions involve averages over the lowest order particle trajectories, that typically cannot be evaluated analytically. In this work, an efficient numerical method for the evaluation of such expressions is reported. It is shown that this method is faster than direct gyrokinetic simulations performed with the Gene and EUTERPE codes. Calculations of the residual value in stellarators are provided for much shorter wavelengths than previously available in the literature. Electrons must be treated kinetically in stellarators because, unlike in tokamaks, kinetic electrons modify the residual value even at long wavelengths. This effect, that had already been predicted theoretically, is confirmed by gyrokinetic simulations.

Edge equilibrium code for tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Xujing; Zakharov, Leonid E.; Drozdov, Vladimir V.

2014-01-15

The edge equilibrium code (EEC) described in this paper is developed for simulations of the near edge plasma using the finite element method. It solves the Grad-Shafranov equation in toroidal coordinate and uses adaptive grids aligned with magnetic field lines. Hermite finite elements are chosen for the numerical scheme. A fast Newton scheme which is the same as implemented in the equilibrium and stability code (ESC) is applied here to adjust the grids.

Full wave simulations of helicon wave losses in the scrape-off-layer of the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Lau, Cornwall; Jaeger, Fred; Berry, Lee; Bertelli, Nicola; Pinsker, Robert

2017-10-01

Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D. Previous modeling using the hot plasma, full wave code AORSA, has shown good agreement with the ray tracing code GENRAY for helicon wave propagation and absorption in the core plasma. AORSA, and a new, reduced finite-element-model show discrepancies between ray tracing and full wave occur in the scrape-off-layer (SOL), especially at high densities. The reduced model is much faster than AORSA, and reproduces most of the important features of the AORSA model. The reduced model also allows for larger parametric scans and for the easy use of arbitrary tokamak geometry. Results of the full wave codes, AORSA and COMSOL, will be shown for helicon wave losses in the SOL are shown for a large range of parameters, such as SOL density profiles, n||, radial and vertical locations of the antenna, and different tokamak vessel geometries. This work was supported by DE-AC05-00OR22725, DE-AC02-09CH11466, and DE-FC02-04ER54698.

Theory-based model for the pedestal, edge stability and ELMs in tokamaks

NASA Astrophysics Data System (ADS)

Pankin, A. Y.; Bateman, G.; Brennan, D. P.; Schnack, D. D.; Snyder, P. B.; Voitsekhovitch, I.; Kritz, A. H.; Janeschitz, G.; Kruger, S.; Onjun, T.; Pacher, G. W.; Pacher, H. D.

2006-04-01

An improved model for triggering edge localized mode (ELM) crashes is developed for use within integrated modelling simulations of the pedestal and ELM cycles at the edge of H-mode tokamak plasmas. The new model is developed by using the BALOO, DCON and ELITE ideal MHD stability codes to derive parametric expressions for the ELM triggering threshold. The whole toroidal mode number spectrum is studied with these codes. The DCON code applies to low mode numbers, while the BALOO code applies to only high mode numbers and the ELITE code applies to intermediate and high mode numbers. The variables used in the parametric stability expressions are the normalized pressure gradient and the parallel current density, which drive ballooning and peeling modes. Two equilibria motivated by DIII-D geometry with different plasma triangularities are studied. It is found that the stable region in the high triangularity discharge covers a much larger region of parameter space than the corresponding stability region in the low triangularity discharge. The new ELM trigger model is used together with a previously developed model for pedestal formation and ELM crashes in the ASTRA integrated modelling code to follow the time evolution of the temperature profiles during ELM cycles. The ELM frequencies obtained in the simulations of low and high triangularity discharges are observed to increase with increasing heating power. There is a transition from second stability to first ballooning mode stability as the heating power is increased in the high triangularity simulations. The results from the ideal MHD stability codes are compared with results from the resistive MHD stability code NIMROD.

A new method for computing the gyrocenter orbit in the tokamak configuration

NASA Astrophysics Data System (ADS)

Xu, Yingfeng

2013-10-01

Gyrokinetic theory is an important tool for studying the long-time behavior of magnetized plasmas in Tokamaks. The gyrocenter trajectory determined by the gyrocenter equations of motion can be computed by using a special kind of the Lie-transform perturbation method. The corresponding Lie-transform called I-transform makes that the transformed equations of motion have the same form as the unperturbed ones. The gyrocenter trajectory in short time is divided into two parts. One is along the unperturbed orbit. The other one, which is related to perturbation, is determined by the I-transform generating vector. The numerical gyrocenter orbit code based on this new method has been developed in the tokamak configuration and benchmarked with the other orbit code in some simple cases. Furthermore, it is clearly demonstrated that this new method for computing gyrocenter orbit is equivalent to the gyrocenter Hamilton equations of motion up to the second order in timestep. The new method can be applied to the gyrokinetic simulation. The gyrocenter orbit of the unperturbed part determined by the equilibrium fields can be computed previously in the gyrokinetic simulation, and the corresponding time consumption is neglectable.

Initial Computations of Vertical Displacement Events with NIMROD

NASA Astrophysics Data System (ADS)

Bunkers, Kyle; Sovinec, C. R.

2014-10-01

Disruptions associated with vertical displacement events (VDEs) have potential for causing considerable physical damage to ITER and other tokamak experiments. We report on initial computations of generic axisymmetric VDEs using the NIMROD code [Sovinec et al., JCP 195, 355 (2004)]. An implicit thin-wall computation has been implemented to couple separate internal and external regions without numerical stability limitations. A simple rectangular cross-section domain generated with the NIMEQ code [Howell and Sovinec, CPC (2014)] modified to use a symmetry condition at the midplane is used to test linear and nonlinear axisymmetric VDE computation. As current in simulated external coils for large- R / a cases is varied, there is a clear n = 0 stability threshold which lies below the decay-index criterion for the current-loop model of a tokamak to model VDEs [Mukhovatov and Shafranov, Nucl. Fusion 11, 605 (1971)]; a scan of wall distance indicates the offset is due to the influence of the conducting wall. Results with a vacuum region surrounding a resistive wall will also be presented. Initial nonlinear computations show large vertical displacement of an intact simulated tokamak. This effort is supported by U.S. Department of Energy Grant DE-FG02-06ER54850.

Multi-region approach to free-boundary three-dimensional tokamak equilibria and resistive wall instabilities

NASA Astrophysics Data System (ADS)

Ferraro, N. M.; Jardin, S. C.; Lao, L. L.; Shephard, M. S.; Zhang, F.

2016-05-01

Free-boundary 3D tokamak equilibria and resistive wall instabilities are calculated using a new resistive wall model in the two-fluid M3D-C1 code. In this model, the resistive wall and surrounding vacuum region are included within the computational domain. This implementation contrasts with the method typically used in fluid codes in which the resistive wall is treated as a boundary condition on the computational domain boundary and has the advantage of maintaining purely local coupling of mesh elements. This new capability is used to simulate perturbed, free-boundary non-axisymmetric equilibria; the linear evolution of resistive wall modes; and the linear and nonlinear evolution of axisymmetric vertical displacement events (VDEs). Calculated growth rates for a resistive wall mode with arbitrary wall thickness are shown to agree well with the analytic theory. Equilibrium and VDE calculations are performed in diverted tokamak geometry, at physically realistic values of dissipation, and with resistive walls of finite width. Simulations of a VDE disruption extend into the current-quench phase, in which the plasma becomes limited by the first wall, and strong currents are observed to flow in the wall, in the SOL, and from the plasma to the wall.

Simulation of current-filament dynamics and relaxation in the Pegasus Spherical Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Bryan, J. B.; Sovinec, C. R.; Bird, T. M.

Nonlinear numerical computation is used to investigate the relaxation of non-axisymmetric current-channels from washer-gun plasma sources into 'tokamak-like' plasmas in the Pegasus toroidal experiment [Eidietis et al. J. Fusion Energy 26, 43 (2007)]. Resistive MHD simulations with the NIMROD code [Sovinec et al. Phys. Plasmas 10(5), 1727-1732 (2003)] utilize ohmic heating, temperature-dependent resistivity, and anisotropic, temperature-dependent thermal conduction corrected for regions of low magnetization to reproduce critical transport effects. Adjacent passes of the simulated current-channel attract and generate strong reversed current sheets that suggest magnetic reconnection. With sufficient injected current, adjacent passes merge periodically, releasing axisymmetric current rings from themore » driven channel. The current rings have not been previously observed in helicity injection for spherical tokamaks, and as such, provide a new phenomenological understanding for filament relaxation in Pegasus. After large-scale poloidal-field reversal, a hollow current profile and significant poloidal flux amplification accumulate over many reconnection cycles.« less

Design of 28 GHz, 200 kW Gyrotron for ECRH Applications

NASA Astrophysics Data System (ADS)

Yadav, Vivek; Singh, Udaybir; Kumar, Nitin; Kumar, Anil; Deorani, S. C.; Sinha, A. K.

2013-01-01

This paper presents the design of 28 GHz, 200 kW gyrotron for Indian TOKAMAK system. The paper reports the designs of interaction cavity, magnetron injection gun and RF window. EGUN code is used for the optimization of electron gun parameters. TE03 mode is selected as the operating mode by using the in-house developed code GCOMS. The simulation and optimization of the cavity parameters are carried out by using the Particle-in-cell, three dimensional (3-D)-electromagnetic simulation code MAGIC. The output power more than 250 kW is achieved.

Edge Equilibrium Code (EEC) For Tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Xujling

2014-02-24

The edge equilibrium code (EEC) described in this paper is developed for simulations of the near edge plasma using the finite element method. It solves the Grad-Shafranov equation in toroidal coordinate and uses adaptive grids aligned with magnetic field lines. Hermite finite elements are chosen for the numerical scheme. A fast Newton scheme which is the same as implemented in the equilibrium and stability code (ESC) is applied here to adjust the grids

DOUBLE code simulations of emissivities of fast neutrals for different plasma observation view-lines of neutral particle analyzers on the COMPASS tokamak

NASA Astrophysics Data System (ADS)

Mitosinkova, K.; Tomes, M.; Stockel, J.; Varju, J.; Stano, M.

2018-03-01

Neutral particle analyzers (NPA) measure line-integrated energy spectra of fast neutral atoms escaping the tokamak plasma, which are a product of charge-exchange (CX) collisions of plasma ions with background neutrals. They can observe variations in the ion temperature T i of non-thermal fast ions created by additional plasma heating. However, the plasma column which a fast atom has to pass through must be sufficiently short in comparison with the fast atom’s mean-free-path. Tokamak COMPASS is currently equipped with one NPA installed at a tangential mid-plane port. This orientation is optimal for observing non-thermal fast ions. However, in this configuration the signal at energies useful for T i derivation is lost in noise due to the too long fast atoms’ trajectories. Thus, a second NPA is planned to be connected for the purpose of measuring T i. We analyzed different possible view-lines (perpendicular mid-plane, tangential mid-plane, and top view) for the second NPA using the DOUBLE Monte-Carlo code and compared the results with the performance of the present NPA with tangential orientation. The DOUBLE code provides fast-atoms’ emissivity functions along the NPA view-line. The position of the median of these emissivity functions is related to the location from where the measured signal originates. Further, we compared the difference between the real central T i used as a DOUBLE code input and the T iCX derived from the exponential decay of simulated energy spectra. The advantages and disadvantages of each NPA location are discussed.

Fully 3D modeling of tokamak vertical displacement events with realistic parameters

NASA Astrophysics Data System (ADS)

Pfefferle, David; Ferraro, Nathaniel; Jardin, Stephen; Bhattacharjee, Amitava

2016-10-01

In this work, we model the complex multi-domain and highly non-linear physics of Vertical Displacement Events (VDEs), one of the most damaging off-normal events in tokamaks, with the implicit 3D extended MHD code M3D-C1. The code has recently acquired the capability to include finite thickness conducting structures within the computational domain. By exploiting the possibility of running a linear 3D calculation on top of a non-linear 2D simulation, we monitor the non-axisymmetric stability and assess the eigen-structure of kink modes as the simulation proceeds. Once a stability boundary is crossed, a fully 3D non-linear calculation is launched for the remainder of the simulation, starting from an earlier time of the 2D run. This procedure, along with adaptive zoning, greatly increases the efficiency of the calculation, and allows to perform VDE simulations with realistic parameters and high resolution. Simulations are being validated with NSTX data where both axisymmetric (toroidally averaged) and non-axisymmetric induced and conductive (halo) currents have been measured. This work is supported by US DOE Grant DE-AC02-09CH11466.

Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, X Q

We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. Withmore » our 4D ({psi}, {theta}, {epsilon}, {mu}) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices.« less

Toward a first-principles integrated simulation of tokamak edge plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang, C S; Klasky, Scott A; Cummings, Julian

2008-01-01

Performance of the ITER is anticipated to be highly sensitive to the edge plasma condition. The edge pedestal in ITER needs to be predicted from an integrated simulation of the necessary firstprinciples, multi-scale physics codes. The mission of the SciDAC Fusion Simulation Project (FSP) Prototype Center for Plasma Edge Simulation (CPES) is to deliver such a code integration framework by (1) building new kinetic codes XGC0 and XGC1, which can simulate the edge pedestal buildup; (2) using and improving the existing MHD codes ELITE, M3D-OMP, M3D-MPP and NIMROD, for study of large-scale edge instabilities called Edge Localized Modes (ELMs); andmore » (3) integrating the codes into a framework using cutting-edge computer science technology. Collaborative effort among physics, computer science, and applied mathematics within CPES has created the first working version of the End-to-end Framework for Fusion Integrated Simulation (EFFIS), which can be used to study the pedestal-ELM cycles.« less

Integrated Tokamak modeling: When physics informs engineering and research planning

NASA Astrophysics Data System (ADS)

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. It discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.

Integrated Tokamak modeling: When physics informs engineering and research planning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poli, Francesca Maria

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

Integrated Tokamak modeling: When physics informs engineering and research planning

DOE PAGES

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

Design of the DEMO Fusion Reactor Following ITER.

PubMed

Garabedian, Paul R; McFadden, Geoffrey B

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task.

Design of the DEMO Fusion Reactor Following ITER

PubMed Central

Garabedian, Paul R.; McFadden, Geoffrey B.

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task. PMID:27504224

Synthetic Microwave Imaging Reflectometry diagnostic using 3D FDTD Simulations

NASA Astrophysics Data System (ADS)

Kruger, Scott; Jenkins, Thomas; Smithe, David; King, Jacob; Nimrod Team Team

2017-10-01

Microwave Imaging Reflectometry (MIR) has become a standard diagnostic for understanding tokamak edge perturbations, including the edge harmonic oscillations in QH mode operation. These long-wavelength perturbations are larger than the normal turbulent fluctuation levels and thus normal analysis of synthetic signals become more difficult. To investigate, we construct a synthetic MIR diagnostic for exploring density fluctuation amplitudes in the tokamak plasma edge by using the three-dimensional, full-wave FDTD code Vorpal. The source microwave beam for the diagnostic is generated and refelected at the cutoff surface that is distorted by 2D density fluctuations in the edge plasma. Synthetic imaging optics at the detector can be used to understand the fluctuation and background density profiles. We apply the diagnostic to understand the fluctuations in edge plasma density during QH-mode activity in the DIII-D tokamak, as modeled by the NIMROD code. This work was funded under DOE Grant Number DE-FC02-08ER54972.

Global two-fluid simulations of geodesic acoustic modes in strongly shaped tight aspect ratio tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robinson, J. R.; Hnat, B.; Thyagaraja, A.

2013-05-15

Following recent observations suggesting the presence of the geodesic acoustic mode (GAM) in ohmically heated discharges in the Mega Amp Spherical Tokamak (MAST) [J. R. Robinson et al., Plasma Phys. Controlled Fusion 54, 105007 (2012)], the behaviour of the GAM is studied numerically using the two fluid, global code CENTORI [P. J. Knight et al. Comput. Phys. Commun. 183, 2346 (2012)]. We examine mode localisation and effects of magnetic geometry, given by aspect ratio, elongation, and safety factor, on the observed frequency of the mode. An excellent agreement between simulations and experimental data is found for simulation plasma parameters matchedmore » to those of MAST. Increasing aspect ratio yields good agreement between the GAM frequency found in the simulations and an analytical result obtained for elongated large aspect ratio plasmas.« less

The accurate particle tracer code

NASA Astrophysics Data System (ADS)

Wang, Yulei; Liu, Jian; Qin, Hong; Yu, Zhi; Yao, Yicun

2017-11-01

The Accurate Particle Tracer (APT) code is designed for systematic large-scale applications of geometric algorithms for particle dynamical simulations. Based on a large variety of advanced geometric algorithms, APT possesses long-term numerical accuracy and stability, which are critical for solving multi-scale and nonlinear problems. To provide a flexible and convenient I/O interface, the libraries of Lua and Hdf5 are used. Following a three-step procedure, users can efficiently extend the libraries of electromagnetic configurations, external non-electromagnetic forces, particle pushers, and initialization approaches by use of the extendible module. APT has been used in simulations of key physical problems, such as runaway electrons in tokamaks and energetic particles in Van Allen belt. As an important realization, the APT-SW version has been successfully distributed on the world's fastest computer, the Sunway TaihuLight supercomputer, by supporting master-slave architecture of Sunway many-core processors. Based on large-scale simulations of a runaway beam under parameters of the ITER tokamak, it is revealed that the magnetic ripple field can disperse the pitch-angle distribution significantly and improve the confinement of energetic runaway beam on the same time.

Three dimensional nonlinear simulations of edge localized modes on the EAST tokamak using BOUT++ code

NASA Astrophysics Data System (ADS)

Liu, Z. X.; Xu, X. Q.; Gao, X.; Xia, T. Y.; Joseph, I.; Meyer, W. H.; Liu, S. C.; Xu, G. S.; Shao, L. M.; Ding, S. Y.; Li, G. Q.; Li, J. G.

2014-09-01

Experimental measurements of edge localized modes (ELMs) observed on the EAST experiment are compared to linear and nonlinear theoretical simulations of peeling-ballooning modes using the BOUT++ code. Simulations predict that the dominant toroidal mode number of the ELM instability becomes larger for lower current, which is consistent with the mode structure captured with visible light using an optical CCD camera. The poloidal mode number of the simulated pressure perturbation shows good agreement with the filamentary structure observed by the camera. The nonlinear simulation is also consistent with the experimentally measured energy loss during an ELM crash and with the radial speed of ELM effluxes measured using a gas puffing imaging diagnostic.

Simulations of toroidal Alfvén eigenmode excited by fast ions on the Experimental Advanced Superconducting Tokamak

NASA Astrophysics Data System (ADS)

Pei, Youbin; Xiang, Nong; Shen, Wei; Hu, Youjun; Todo, Y.; Zhou, Deng; Huang, Juan

2018-05-01

Kinetic-MagnetoHydroDynamic (MHD) hybrid simulations are carried out to study fast ion driven toroidal Alfvén eigenmodes (TAEs) on the Experimental Advanced Superconducting Tokamak (EAST). The first part of this article presents the linear benchmark between two kinetic-MHD codes, namely MEGA and M3D-K, based on a realistic EAST equilibrium. Parameter scans show that the frequency and the growth rate of the TAE given by the two codes agree with each other. The second part of this article discusses the resonance interaction between the TAE and fast ions simulated by the MEGA code. The results show that the TAE exchanges energy with the co-current passing particles with the parallel velocity |v∥ | ≈VA 0/3 or |v∥ | ≈VA 0/5 , where VA 0 is the Alfvén speed on the magnetic axis. The TAE destabilized by the counter-current passing ions is also analyzed and found to have a much smaller growth rate than the co-current ions driven TAE. One of the reasons for this is found to be that the overlapping region of the TAE spatial location and the counter-current ion orbits is narrow, and thus the wave-particle energy exchange is not efficient.

Simulations of vertical disruptions with VDE code: Hiro and Evans currents

NASA Astrophysics Data System (ADS)

Li, Xujing; Di Hu Team; Leonid Zakharov Team; Galkin Team

2014-10-01

The recently created numerical code VDE for simulations of vertical instability in tokamaks is presented. The numerical scheme uses the Tokamak MHD model, where the plasma inertia is replaced by the friction force, and an adaptive grid numerical scheme. The code reproduces well the surface currents generated at the plasma boundary by the instability. Five regimes of the vertical instability are presented: (1) Vertical instability in a given plasma shaping field without a wall; (2) The same with a wall and magnetic flux ΔΨ|plX< ΔΨ|Xwall(where X corresponds to the X-point of a separatrix); (3) The same with a wall and magnetic flux ΔΨ|plX> ΔΨ|Xwall; (4) Vertical instability without a wall with a tile surface at the plasma path; (5) The same in the presence of a wall and a tile surface. The generation of negative Hiro currents along the tile surface, predicted earlier by the theory and measured on EAST in 2012, is well-reproduced by simulations. In addition, the instability generates the force-free Evans currents at the free plasma surface. The new pattern of reconnection of the plasma with the vacuum magnetic field is discovered. This work is supported by US DoE Contract No. DE-AC02-09-CH11466.

Multi-region approach to free-boundary three-dimensional tokamak equilibria and resistive wall instabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferraro, N. M.; Jardin, S. C.; Lao, L. L.

Free-boundary 3D tokamak equilibria and resistive wall instabilities are calculated using a new resistive wall model in the two-fluid M3D-C1 code. In this model, the resistive wall and surround- ing vacuum region are included within the computational domain. Our implementation contrasts with the method typically used in fluid codes in which the resistive wall is treated as a boundary condition on the computational domain boundary and has the advantage of maintaining purely local coupling of mesh elements. We use this new capability to simulate perturbed, free-boundary non- axisymmetric equilibria; the linear evolution of resistive wall modes; and the linear andmore » nonlinear evolution of axisymmetric vertical displacement events (VDEs). Calculated growth rates for a resistive wall mode with arbitrary wall thickness are shown to agree well with the analytic theory. Equilibrium and VDE calculations are performed in diverted tokamak geometry, at physically real- istic values of dissipation, and with resistive walls of finite width. Simulations of a VDE disruption extend into the current-quench phase, in which the plasma becomes limited by the first wall, and strong currents are observed to flow in the wall, in the SOL, and from the plasma to the wall.« less

Multi-region approach to free-boundary three-dimensional tokamak equilibria and resistive wall instabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferraro, N. M., E-mail: nferraro@pppl.gov; Lao, L. L.; Jardin, S. C.

Free-boundary 3D tokamak equilibria and resistive wall instabilities are calculated using a new resistive wall model in the two-fluid M3D-C1 code. In this model, the resistive wall and surrounding vacuum region are included within the computational domain. This implementation contrasts with the method typically used in fluid codes in which the resistive wall is treated as a boundary condition on the computational domain boundary and has the advantage of maintaining purely local coupling of mesh elements. This new capability is used to simulate perturbed, free-boundary non-axisymmetric equilibria; the linear evolution of resistive wall modes; and the linear and nonlinear evolutionmore » of axisymmetric vertical displacement events (VDEs). Calculated growth rates for a resistive wall mode with arbitrary wall thickness are shown to agree well with the analytic theory. Equilibrium and VDE calculations are performed in diverted tokamak geometry, at physically realistic values of dissipation, and with resistive walls of finite width. Simulations of a VDE disruption extend into the current-quench phase, in which the plasma becomes limited by the first wall, and strong currents are observed to flow in the wall, in the SOL, and from the plasma to the wall.« less

Multi-region approach to free-boundary three-dimensional tokamak equilibria and resistive wall instabilities

DOE PAGES

Ferraro, N. M.; Jardin, S. C.; Lao, L. L.; ...

2016-05-20

Free-boundary 3D tokamak equilibria and resistive wall instabilities are calculated using a new resistive wall model in the two-fluid M3D-C1 code. In this model, the resistive wall and surround- ing vacuum region are included within the computational domain. Our implementation contrasts with the method typically used in fluid codes in which the resistive wall is treated as a boundary condition on the computational domain boundary and has the advantage of maintaining purely local coupling of mesh elements. We use this new capability to simulate perturbed, free-boundary non- axisymmetric equilibria; the linear evolution of resistive wall modes; and the linear andmore » nonlinear evolution of axisymmetric vertical displacement events (VDEs). Calculated growth rates for a resistive wall mode with arbitrary wall thickness are shown to agree well with the analytic theory. Equilibrium and VDE calculations are performed in diverted tokamak geometry, at physically real- istic values of dissipation, and with resistive walls of finite width. Simulations of a VDE disruption extend into the current-quench phase, in which the plasma becomes limited by the first wall, and strong currents are observed to flow in the wall, in the SOL, and from the plasma to the wall.« less

On the breakdown modes and parameter space of Ohmic Tokamak startup

NASA Astrophysics Data System (ADS)

Peng, Yanli; Jiang, Wei; Zhang, Ya; Hu, Xiwei; Zhuang, Ge; Innocenti, Maria; Lapenta, Giovanni

2017-10-01

Tokamak plasma has to be hot. The process of turning the initial dilute neutral hydrogen gas at room temperature into fully ionized plasma is called tokamak startup. Even with over 40 years of research, the parameter ranges for the successful startup still aren't determined by numerical simulations but by trial and errors. However, in recent years it has drawn much attention due to one of the challenges faced by ITER: the maximum electric field for startup can't exceed 0.3 V/m, which makes the parameter range for successful startup narrower. Besides, this physical mechanism is far from being understood either theoretically or numerically. In this work, we have simulated the plasma breakdown phase driven by pure Ohmic heating using a particle-in-cell/Monte Carlo code, with the aim of giving a predictive parameter range for most tokamaks, even for ITER. We have found three situations during the discharge, as a function of the initial parameters: no breakdown, breakdown and runaway. Moreover, breakdown delay and volt-second consumption under different initial conditions are evaluated. In addition, we have simulated breakdown on ITER and confirmed that when the electric field is 0.3 V/m, the optimal pre-filling pressure is 0.001 Pa, which is in good agreement with ITER's design.

Structure of the classical scrape-off layer of a tokamak

NASA Astrophysics Data System (ADS)

Rozhansky, V.; Kaveeva, E.; Senichenkov, I.; Vekshina, E.

2018-03-01

The structure of the scrape-off layer (SOL) of a tokamak with little or no turbulent transport is analyzed. The analytical estimates of the density and electron temperature fall-off lengths of the SOL are put forward. It is demonstrated that the SOL width could be of the order of the ion poloidal gyroradius, as suggested in Goldston (2012 Nuclear Fusion 52 013009). The analytical results are supported by the results of the 2D simulations of the edge plasma with reduced transport coefficients performed by SOLPS-ITER transport code.

Initial Simulations of RF Waves in Hot Plasmas Using the FullWave Code

NASA Astrophysics Data System (ADS)

Zhao, Liangji; Svidzinski, Vladimir; Spencer, Andrew; Kim, Jin-Soo

2017-10-01

FullWave is a simulation tool that models RF fields in hot inhomogeneous magnetized plasmas. The wave equations with linearized hot plasma dielectric response are solved in configuration space on adaptive cloud of computational points. The nonlocal hot plasma dielectric response is formulated by calculating the plasma conductivity kernel based on the solution of the linearized Vlasov equation in inhomogeneous magnetic field. In an rf field, the hot plasma dielectric response is limited to the distance of a few particles' Larmor radii, near the magnetic field line passing through the test point. The localization of the hot plasma dielectric response results in a sparse matrix of the problem thus significantly reduces the size of the problem and makes the simulations faster. We will present the initial results of modeling of rf waves using the Fullwave code, including calculation of nonlocal conductivity kernel in 2D Tokamak geometry; the interpolation of conductivity kernel from test points to adaptive cloud of computational points; and the results of self-consistent simulations of 2D rf fields using calculated hot plasma conductivity kernel in a tokamak plasma with reduced parameters. Work supported by the US DOE ``SBIR program.

Edge-relevant plasma simulations with the continuum code COGENT

NASA Astrophysics Data System (ADS)

Dorf, M.; Dorr, M.; Ghosh, D.; Hittinger, J.; Rognlien, T.; Cohen, R.; Lee, W.; Schwartz, P.

2016-10-01

We describe recent advances in cross-separatrix and other edge-relevant plasma simulations with COGENT, a continuum gyro-kinetic code being developed by the Edge Simulation Laboratory (ESL) collaboration. The distinguishing feature of the COGENT code is its high-order finite-volume discretization methods, which employ arbitrary mapped multiblock grid technology (nearly field-aligned on blocks) to handle the complexity of tokamak divertor geometry with high accuracy. This paper discusses the 4D (axisymmetric) electrostatic version of the code, and the presented topics include: (a) initial simulations with kinetic electrons and development of reduced fluid models; (b) development and application of implicit-explicit (IMEX) time integration schemes; and (c) conservative modeling of drift-waves and the universal instability. Work performed for USDOE, at LLNL under contract DE-AC52-07NA27344 and at LBNL under contract DE-AC02-05CH11231.

Some Aspects of Advanced Tokamak Modeling in DIII-D

NASA Astrophysics Data System (ADS)

St John, H. E.; Petty, C. C.; Murakami, M.; Kinsey, J. E.

2000-10-01

We extend previous work(M. Murakami, et al., General Atomics Report GA-A23310 (1999).) done on time dependent DIII-D advanced tokamak simulations by introducing theoretical confinement models rather than relying on power balance derived transport coefficients. We explore using NBCD and off axis ECCD together with a self-consistent aligned bootstrap current, driven by the internal transport barrier dynamics generated with the GLF23 confinement model, to shape the hollow current profile and to maintain MHD stable conditions. Our theoretical modeling approach uses measured DIII-D initial conditions to start off the simulations in a smooth consistent manner. This mitigates the troublesome long lived perturbations in the ohmic current profile that is normally caused by inconsistent initial data. To achieve this goal our simulation uses a sequence of time dependent eqdsks generated autonomously by the EFIT MHD equilibrium code in analyzing experimental data to supply the history for the simulation.

Simulation of MST tokamak discharges with resonant magnetic perturbations

NASA Astrophysics Data System (ADS)

Cornille, B. S.; Sovinec, C. R.; Chapman, B. E.; Dubois, A.; McCollam, K. J.; Munaretto, S.

2016-10-01

Nonlinear MHD modeling of MST tokamak plasmas with an applied resonant magnetic perturbation (RMP) reveals degradation of flux surfaces that may account for the experimentally observed suppression of runaway electrons with the RMP. Runaway electrons are routinely generated in MST tokamak discharges with low plasma density. When an m = 3 RMP is applied these electrons are strongly suppressed, while an m = 1 RMP of comparable amplitude has little effect. The computations are performed using the NIMROD code and use reconstructed equilibrium states of MST tokamak plasmas with q (0) < 1 and q (a) = 2.2 . Linear computations show that the (1 , 1) -kink and (2 , 2) -tearing modes are unstable, and nonlinear simulations produce sawtoothing with a period of approximately 0.5 ms, which is comparable to the period of MHD activity observed experimentally. Adding an m = 3 RMP in the computation degrades flux surfaces in the outer region of the plasma, while no degradation occurs with an m = 1 RMP. The outer flux surface degradation with the m = 3 RMP, combined with the sawtooth-induced distortion of flux surfaces in the core, may account for the observed suppression of runaway electrons. Work supported by DOE Grant DE-FC02-08ER54975.

Simulations of 4D edge transport and dynamics using the TEMPEST gyro-kinetic code

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Cohen, B. I.; Cohen, R. H.; Dorr, M. R.; Hittinger, J. A. F.; Kerbel, G. D.; Nevins, W. M.; Xiong, Z.; Xu, X. Q.

2006-10-01

Simulation results are presented for tokamak edge plasmas with a focus on the 4D (2r,2v) option of the TEMPEST continuum gyro-kinetic code. A detailed description of a variety of kinetic simulations is reported, including neoclassical radial transport from Coulomb collisions, electric field generation, dynamic response to perturbations by geodesic acoustic modes, and parallel transport on open magnetic-field lines. Comparison is made between the characteristics of the plasma solutions on closed and open magnetic-field line regions separated by a magnetic separatrix, and simple physical models are used to qualitatively explain the differences observed in mean flow and electric-field generation. The status of extending the simulations to 5D turbulence will be summarized. The code structure used in this ongoing project is also briefly described, together with future plans.

Gyrokinetic Particle Simulation of Turbulent Transport in Burning Plasmas (GPS - TTBP) Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chame, Jacqueline

2011-05-27

The goal of this project is the development of the Gyrokinetic Toroidal Code (GTC) Framework and its applications to problems related to the physics of turbulence and turbulent transport in tokamaks,. The project involves physics studies, code development, noise effect mitigation, supporting computer science efforts, diagnostics and advanced visualizations, verification and validation. Its main scientific themes are mesoscale dynamics and non-locality effects on transport, the physics of secondary structures such as zonal flows, and strongly coherent wave-particle interaction phenomena at magnetic precession resonances. Special emphasis is placed on the implications of these themes for rho-star and current scalings and formore » the turbulent transport of momentum. GTC-TTBP also explores applications to electron thermal transport, particle transport; ITB formation and cross-cuts such as edge-core coupling, interaction of energetic particles with turbulence and neoclassical tearing mode trigger dynamics. Code development focuses on major initiatives in the development of full-f formulations and the capacity to simulate flux-driven transport. In addition to the full-f -formulation, the project includes the development of numerical collision models and methods for coarse graining in phase space. Verification is pursued by linear stability study comparisons with the FULL and HD7 codes and by benchmarking with the GKV, GYSELA and other gyrokinetic simulation codes. Validation of gyrokinetic models of ion and electron thermal transport is pursed by systematic stressing comparisons with fluctuation and transport data from the DIII-D and NSTX tokamaks. The physics and code development research programs are supported by complementary efforts in computer sciences, high performance computing, and data management.« less

Comparative modelling of lower hybrid current drive with two launcher designs in the Tore Supra tokamak

NASA Astrophysics Data System (ADS)

Nilsson, E.; Decker, J.; Peysson, Y.; Artaud, J.-F.; Ekedahl, A.; Hillairet, J.; Aniel, T.; Basiuk, V.; Goniche, M.; Imbeaux, F.; Mazon, D.; Sharma, P.

2013-08-01

Fully non-inductive operation with lower hybrid current drive (LHCD) in the Tore Supra tokamak is achieved using either a fully active multijunction (FAM) launcher or a more recent ITER-relevant passive active multijunction (PAM) launcher, or both launchers simultaneously. While both antennas show comparable experimental efficiencies, the analysis of stability properties in long discharges suggest different current profiles. We present comparative modelling of LHCD with the two different launchers to characterize the effect of the respective antenna spectra on the driven current profile. The interpretative modelling of LHCD is carried out using a chain of codes calculating, respectively, the global discharge evolution (tokamak simulator METIS), the spectrum at the antenna mouth (LH coupling code ALOHA), the LH wave propagation (ray-tracing code C3PO), and the distribution function (3D Fokker-Planck code LUKE). Essential aspects of the fast electron dynamics in time, space and energy are obtained from hard x-ray measurements of fast electron bremsstrahlung emission using a dedicated tomographic system. LHCD simulations are validated by systematic comparisons between these experimental measurements and the reconstructed signal calculated by the code R5X2 from the LUKE electron distribution. An excellent agreement is obtained in the presence of strong Landau damping (found under low density and high-power conditions in Tore Supra) for which the ray-tracing model is valid for modelling the LH wave propagation. Two aspects of the antenna spectra are found to have a significant effect on LHCD. First, the driven current is found to be proportional to the directivity, which depends upon the respective weight of the main positive and main negative lobes and is particularly sensitive to the density in front of the antenna. Second, the position of the main negative lobe in the spectrum is different for the two launchers. As this lobe drives a counter-current, the resulting driven current profile is also different for the FAM and PAM launchers.

Four-Dimensional Continuum Gyrokinetic Code: Neoclassical Simulation of Fusion Edge Plasmas

NASA Astrophysics Data System (ADS)

Xu, X. Q.

2005-10-01

We are developing a continuum gyrokinetic code, TEMPEST, to simulate edge plasmas. Our code represents velocity space via a grid in equilibrium energy and magnetic moment variables, and configuration space via poloidal magnetic flux and poloidal angle. The geometry is that of a fully diverted tokamak (single or double null) and so includes boundary conditions for both closed magnetic flux surfaces and open field lines. The 4-dimensional code includes kinetic electrons and ions, and electrostatic field-solver options, and simulates neoclassical transport. The present implementation is a Method of Lines approach where spatial finite-differences (higher order upwinding) and implicit time advancement are used. We present results of initial verification and validation studies: transition from collisional to collisionless limits of parallel end-loss in the scrape-off layer, self-consistent electric field, and the effect of the real X-point geometry and edge plasma conditions on the standard neoclassical theory, including a comparison of our 4D code with other kinetic neoclassical codes and experiments.

The effects of resonant magnetic perturbations on fast ion confinement in the Mega Amp Spherical Tokamak

NASA Astrophysics Data System (ADS)

McClements, K. G.; Akers, R. J.; Boeglin, W. U.; Cecconello, M.; Keeling, D.; Jones, O. M.; Kirk, A.; Klimek, I.; Perez, R. V.; Shinohara, K.; Tani, K.

2015-07-01

The effects of resonant magnetic perturbations (RMPs) on the confinement of energetic (neutral beam) ions in the Mega Amp Spherical Tokamak (MAST) are assessed experimentally using measurements of neutrons, fusion protons and fast ion Dα (FIDA) light emission. In single null-diverted (SND) MAST pulses with relatively low plasma current (400 kA), the total neutron emission dropped by approximately a factor of two when RMPs with toroidal mode number n = 3 were applied. The measured neutron rate during RMPs was much lower than that calculated using the TRANSP plasma simulation code, even when non-classical (but axisymmetric) ad hoc fast ion transport was taken into account in the latter. Sharp drops in spatially-resolved neutron rates, fusion proton rates and FIDA emission were also observed. First principles-based simulations of RMP-induced fast ion transport in MAST, using the F3D-OFMC code, show similar losses for two alternative representations of the MAST first wall, with and without full orbit effects taken into account; for n = 6 RMPs in a 600 kA plasma, the additional loss of beam power due to the RMPs was found in the simulations to be approximately 11%.

The accurate particle tracer code

DOE PAGES

Wang, Yulei; Liu, Jian; Qin, Hong; ...

2017-07-20

The Accurate Particle Tracer (APT) code is designed for systematic large-scale applications of geometric algorithms for particle dynamical simulations. Based on a large variety of advanced geometric algorithms, APT possesses long-term numerical accuracy and stability, which are critical for solving multi-scale and nonlinear problems. To provide a flexible and convenient I/O interface, the libraries of Lua and Hdf5 are used. Following a three-step procedure, users can efficiently extend the libraries of electromagnetic configurations, external non-electromagnetic forces, particle pushers, and initialization approaches by use of the extendible module. APT has been used in simulations of key physical problems, such as runawaymore » electrons in tokamaks and energetic particles in Van Allen belt. As an important realization, the APT-SW version has been successfully distributed on the world’s fastest computer, the Sunway TaihuLight supercomputer, by supporting master–slave architecture of Sunway many-core processors. Here, based on large-scale simulations of a runaway beam under parameters of the ITER tokamak, it is revealed that the magnetic ripple field can disperse the pitch-angle distribution significantly and improve the confinement of energetic runaway beam on the same time.« less

The accurate particle tracer code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Yulei; Liu, Jian; Qin, Hong

The Accurate Particle Tracer (APT) code is designed for systematic large-scale applications of geometric algorithms for particle dynamical simulations. Based on a large variety of advanced geometric algorithms, APT possesses long-term numerical accuracy and stability, which are critical for solving multi-scale and nonlinear problems. To provide a flexible and convenient I/O interface, the libraries of Lua and Hdf5 are used. Following a three-step procedure, users can efficiently extend the libraries of electromagnetic configurations, external non-electromagnetic forces, particle pushers, and initialization approaches by use of the extendible module. APT has been used in simulations of key physical problems, such as runawaymore » electrons in tokamaks and energetic particles in Van Allen belt. As an important realization, the APT-SW version has been successfully distributed on the world’s fastest computer, the Sunway TaihuLight supercomputer, by supporting master–slave architecture of Sunway many-core processors. Here, based on large-scale simulations of a runaway beam under parameters of the ITER tokamak, it is revealed that the magnetic ripple field can disperse the pitch-angle distribution significantly and improve the confinement of energetic runaway beam on the same time.« less

Coupled Kinetic-MHD Simulations of Divertor Heat Load with ELM Perturbations

NASA Astrophysics Data System (ADS)

Cummings, Julian; Chang, C. S.; Park, Gunyoung; Sugiyama, Linda; Pankin, Alexei; Klasky, Scott; Podhorszki, Norbert; Docan, Ciprian; Parashar, Manish

2010-11-01

The effect of Type-I ELM activity on divertor plate heat load is a key component of the DOE OFES Joint Research Target milestones for this year. In this talk, we present simulations of kinetic edge physics, ELM activity, and the associated divertor heat loads in which we couple the discrete guiding-center neoclassical transport code XGC0 with the nonlinear extended MHD code M3D using the End-to-end Framework for Fusion Integrated Simulations, or EFFIS. In these coupled simulations, the kinetic code and the MHD code run concurrently on the same massively parallel platform and periodic data exchanges are performed using a memory-to-memory coupling technology provided by EFFIS. The M3D code models the fast ELM event and sends frequent updates of the magnetic field perturbations and electrostatic potential to XGC0, which in turn tracks particle dynamics under the influence of these perturbations and collects divertor particle and energy flux statistics. We describe here how EFFIS technologies facilitate these coupled simulations and discuss results for DIII-D, NSTX and Alcator C-Mod tokamak discharges.

Likelihood of Alfvénic instability bifurcation in experiments

NASA Astrophysics Data System (ADS)

Duarte, Vinicius; Gorelenkov, Nikolai; Schneller, Mirjam; Fredrickson, Eric; Berk, Herbert; Canal, Gustavo; Heidbrink, William; Kaye, Stanley; Podesta, Mario; van Zeeland, Michael; Wang, Weixing

2017-10-01

We apply a criterion for the likely nature of fast ion redistribution in tokamaks to be in the convective or diffusive nonlinear regimes. The criterion, which is shown to be rather sensitive to the relative strength of collisional or micro-turbulent scattering and drag processes, ultimately translates into a condition for the applicability of reduced quasilinear modeling for realistic tokamak eigenmodes scenarios. The criterion is tested and validated against different machines, where the chirping mode behavior is shown to be in accord with the model. It has been found that the anomalous fast ion transport is a likely mediator of the bifurcation between the fixed-frequency mode behavior and rapid chirping in tokamaks. In addition, micro-turbulence appears to resolve the disparity with respect to the ubiquitous chirping observation in spherical tokamaks and its rarer occurrence in conventional tokamaks. In NSTX, the tendency for chirping is further studied in terms of the beam beta and the plasma rotation shear. For more accurate quantitative assessment, numerical simulations of the effects of electrostatic ion temperature gradient turbulence on chirping are presently being pursued using the GTS code.

Experimental Validation Plan for the Xolotl Plasma-Facing Component Simulator Using Tokamak Sample Exposures

NASA Astrophysics Data System (ADS)

Chan, V. S.; Wong, C. P. C.; McLean, A. G.; Luo, G. N.; Wirth, B. D.

2013-10-01

The Xolotl code under development by PSI-SciDAC will enhance predictive modeling capability of plasma-facing materials under burning plasma conditions. The availability and application of experimental data to compare to code-calculated observables are key requirements to validate the breadth and content of physics included in the model and ultimately gain confidence in its results. A dedicated effort has been in progress to collect and organize a) a database of relevant experiments and their publications as previously carried out at sample exposure facilities in US and Asian tokamaks (e.g., DIII-D DiMES, and EAST MAPES), b) diagnostic and surface analysis capabilities available at each device, and c) requirements for future experiments with code validation in mind. The content of this evolving database will serve as a significant resource for the plasma-material interaction (PMI) community. Work supported in part by the US Department of Energy under GA-DE-SC0008698, DE-AC52-07NA27344 and DE-AC05-00OR22725.

ETF system code: composition and applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reid, R.L.; Wu, K.F.

1980-01-01

A computer code has been developed for application to ETF tokamak system and conceptual design studies. The code determines cost, performance, configuration, and technology requirements as a function of tokamak parameters. The ETF code is structured in a modular fashion in order to allow independent modeling of each major tokamak component. The primary benefit of modularization is that it allows updating of a component module, such as the TF coil module, without disturbing the remainder of the system code as long as the input/output to the modules remains unchanged. The modules may be run independently to perform specific design studies,more » such as determining the effect of allowable strain on TF coil structural requirements, or the modules may be executed together as a system to determine global effects, such as defining the impact of aspect ratio on the entire tokamak system.« less

Simulation of ion-temperature-gradient turbulence in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cohen, B I; Dimits, A M; Kim, C

Results are presented from nonlinear gyrokinetic simulations of toroidal ion temperature gradient (ITG) turbulence and transport. The gyrokinetic simulations are found to yield values of the thermal diffusivity significantly lower than gyrofluid or IFS-PPPL-model predictions. A new phenomenon of nonlinear effective critical gradients larger than the linear instability threshold gradients is observed, and is associated with undamped flux-surface-averaged shear flows. The nonlinear gyrokineic codes have passed extensive validity tests which include comparison against independent linear calculations, a series of nonlinear convergence tests, and a comparison between two independent nonlinear gyrokinetic codes. Our most realistic simulations to date have actual reconstructedmore » equilibria from experiments and a model for dilution by impurity and beam ions. These simulations highlight the need for still more physics to be included in the simulations« less

TEMPEST simulations of the plasma transport in a single-null tokamak geometry

NASA Astrophysics Data System (ADS)

Xu, X. Q.; Bodi, K.; Cohen, R. H.; Krasheninnikov, S.; Rognlien, T. D.

2010-06-01

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. To study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. A series of TEMPEST simulations were conducted to investigate the transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. We also show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arnold H. Kritz

PTRANSP, which is the predictive version of the TRANSP code, was developed in a collaborative effort involving the Princeton Plasma Physics Laboratory, General Atomics Corporation, Lawrence Livermore National Laboratory, and Lehigh University. The PTRANSP/TRANSP suite of codes is the premier integrated tokamak modeling software in the United States. A production service for PTRANSP/TRANSP simulations is maintained at the Princeton Plasma Physics Laboratory; the server has a simple command line client interface and is subscribed to by about 100 researchers from tokamak projects in the US, Europe, and Asia. This service produced nearly 13000 PTRANSP/TRANSP simulations in the four year periodmore » FY 2005 through FY 2008. Major archives of TRANSP results are maintained at PPPL, MIT, General Atomics, and JET. Recent utilization, counting experimental analysis simulations as well as predictive simulations, more than doubled from slightly over 2000 simulations per year in FY 2005 and FY 2006 to over 4300 simulations per year in FY 2007 and FY 2008. PTRANSP predictive simulations applied to ITER increased eight fold from 30 simulations per year in FY 2005 and FY 2006 to 240 simulations per year in FY 2007 and FY 2008, accounting for more than half of combined PTRANSP/TRANSP service CPU resource utilization in FY 2008. PTRANSP studies focused on ITER played a key role in journal articles. Examples of validation studies carried out for momentum transport in PTRANSP simulations were presented at the 2008 IAEA conference. The increase in number of PTRANSP simulations has continued (more than 7000 TRANSP/PTRANSP simulations in 2010) and results of PTRANSP simulations appear in conference proceedings, for example the 2010 IAEA conference, and in peer reviewed papers. PTRANSP provides a bridge to the Fusion Simulation Program (FSP) and to the future of integrated modeling. Through years of widespread usage, each of the many parts of the PTRANSP suite of codes has been thoroughly validated against experimental data and benchmarked against other codes. At the same time, architectural modernizations are improving the modularity of the PTRANSP code base. The NUBEAM neutral beam and fusion products fast ion model, the Plasma State data repository (developed originally in the SWIM SciDAC project and adapted for use in PTRANSP), and other components are already shared with the SWIM, FACETS, and CPES SciDAC FSP prototype projects. Thus, the PTRANSP code is already serving as a bridge between our present integrated modeling capability and future capability. As the Fusion Simulation Program builds toward the facility currently available in the PTRANSP suite of codes, early versions of the FSP core plasma model will need to be benchmarked against the PTRANSP simulations. This will be necessary to build user confidence in FSP, but this benchmarking can only be done if PTRANSP itself is maintained and developed.« less

Modeling MHD Equilibrium and Dynamics with Non-Axisymmetric Resistive Walls in LTX and HBT-EP

NASA Astrophysics Data System (ADS)

Hansen, C.; Levesque, J.; Boyle, D. P.; Hughes, P.

2017-10-01

In experimental magnetized plasmas, currents in the first wall, vacuum vessel, and other conducting structures can have a strong influence on plasma shape and dynamics. These effects are complicated by the 3D nature of these structures, which dictate available current paths. Results from simulations to study the effect of external currents on plasmas in two different experiments will be presented: 1) The arbitrary geometry, 3D extended MHD code PSI-Tet is applied to study linear and non-linear plasma dynamics in the High Beta Tokamak (HBT-EP) focusing on toroidal asymmetries in the adjustable conducting wall. 2) Equilibrium reconstructions of the Lithium Tokamak eXperiment (LTX) in the presence of non-axisymmetric eddy currents. An axisymmetric model is used to reconstruct the plasma equilibrium, using the PSI-Tri code, along with a set of fixed 3D eddy current distributions in the first wall and vacuum vessel [C. Hansen et al., PoP Apr. 2017]. Simulations of detailed experimental geometries are enabled by use of the PSI-Tet code, which employs a high order finite element method on unstructured tetrahedral grids that are generated directly from CAD models. Further development of PSI-Tet and PSI-Tri will also be presented. This work supported by US DOE contract DE-SC0016256.

Double-null divertor configuration discharge and disruptive heat flux simulation using TSC on EAST

NASA Astrophysics Data System (ADS)

Bo, SHI; Jinhong, YANG; Cheng, YANG; Desheng, CHENG; Hui, WANG; Hui, ZHANG; Haifei, DENG; Junli, QI; Xianzu, GONG; Weihua, WANG

2018-07-01

The tokamak simulation code (TSC) is employed to simulate the complete evolution of a disruptive discharge in the experimental advanced superconducting tokamak. The multiplication factor of the anomalous transport coefficient was adjusted to model the major disruptive discharge with double-null divertor configuration based on shot 61 916. The real-time feed-back control system for the plasma displacement was employed. Modeling results of the evolution of the poloidal field coil currents, the plasma current, the major radius, the plasma configuration all show agreement with experimental measurements. Results from the simulation show that during disruption, heat flux about 8 MW m‑2 flows to the upper divertor target plate and about 6 MW m‑2 flows to the lower divertor target plate. Computations predict that different amounts of heat fluxes on the divertor target plate could result by adjusting the multiplication factor of the anomalous transport coefficient. This shows that TSC has high flexibility and predictability.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zakharov, Leonic E.; Li, Xujing

This paper formulates the Tokamak Magneto-Hydrodynamics (TMHD), initially outlined by X. Li and L.E. Zakharov [Plasma Science and Technology, accepted, ID:2013-257 (2013)] for proper simulations of macroscopic plasma dynamics. The simplest set of magneto-hydrodynamics equations, sufficient for disruption modeling and extendable to more refined physics, is explained in detail. First, the TMHD introduces to 3-D simulations the Reference Magnetic Coordinates (RMC), which are aligned with the magnetic field in the best possible way. The numerical implementation of RMC is adaptive grids. Being consistent with the high anisotropy of the tokamak plasma, RMC allow simulations at realistic, very high plasma electricmore » conductivity. Second, the TMHD splits the equation of motion into an equilibrium equation and the plasma advancing equation. This resolves the 4 decade old problem of Courant limitations of the time step in existing, plasma inertia driven numerical codes. The splitting allows disruption simulations on a relatively slow time scale in comparison with the fast time of ideal MHD instabilities. A new, efficient numerical scheme is proposed for TMHD.« less

Continuum kinetic modeling of the tokamak plasma edge

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dorf, M. A.; Dorr, M. R.; Hittinger, J. A.

2016-05-15

The first 4D (axisymmetric) high-order continuum gyrokinetic transport simulations that span the magnetic separatrix of a tokamak are presented. The modeling is performed with the COGENT code, which is distinguished by fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasma transport and the complex X-point divertor geometry with high accuracy. The calculations take into account the effects of fully nonlinear Fokker-Plank collisions, electrostatic potential variations, and anomalous radial transport. Topics discussed include: (a) ion orbit loss and the associated toroidal rotation and (b) edge plasma relaxation in the presence of anomalous radial transport.

Simulations of Neon Pellets for Plasma Disruption Mitigation in Tokamaks

NASA Astrophysics Data System (ADS)

Bosviel, Nicolas; Samulyak, Roman; Parks, Paul

2017-10-01

Numerical studies of the ablation of neon pellets in tokamaks in the plasma disruption mitigation parameter space have been performed using a time-dependent pellet ablation model based on the front tracking code FronTier-MHD. The main features of the model include the explicit tracking of the solid pellet/ablated gas interface, a self-consistent evolving potential distribution in the ablation cloud, JxB forces, atomic processes, and an improved electrical conductivity model. The equation of state model accounts for atomic processes in the ablation cloud as well as deviations from the ideal gas law in the dense, cold layers of neon gas near the pellet surface. Simulations predict processes in the ablation cloud and pellet ablation rates and address the sensitivity of pellet ablation processes to details of physics models, in particular the equation of state.

Comparisons of 'Identical' Simulations by the Eulerian Gyrokinetic Codes GS2 and GYRO

NASA Astrophysics Data System (ADS)

Bravenec, R. V.; Ross, D. W.; Candy, J.; Dorland, W.; McKee, G. R.

2003-10-01

A major goal of the fusion program is to be able to predict tokamak transport from first-principles theory. To this end, the Eulerian gyrokinetic code GS2 was developed years ago and continues to be improved [1]. Recently, the Eulerian code GYRO was developed [2]. These codes are not subject to the statistical noise inherent to particle-in-cell (PIC) codes, and have been very successful in treating electromagnetic fluctuations. GS2 is fully spectral in the radial coordinate while GYRO uses finite-differences and ``banded" spectral schemes. To gain confidence in nonlinear simulations of experiment with these codes, ``apples-to-apples" comparisons (identical profile inputs, flux-tube geometry, two species, etc.) are first performed. We report on a series of linear and nonlinear comparisons (with overall agreement) including kinetic electrons, collisions, and shaped flux surfaces. We also compare nonlinear simulations of a DIII-D discharge to measurements of not only the fluxes but also the turbulence parameters. [1] F. Jenko, et al., Phys. Plasmas 7, 1904 (2000) and refs. therein. [2] J. Candy, J. Comput. Phys. 186, 545 (2003).

NIMROD: A computational laboratory for studying nonlinear fusion magnetohydrodynamics

NASA Astrophysics Data System (ADS)

Sovinec, C. R.; Gianakon, T. A.; Held, E. D.; Kruger, S. E.; Schnack, D. D.

2003-05-01

Nonlinear numerical studies of macroscopic modes in a variety of magnetic fusion experiments are made possible by the flexible high-order accurate spatial representation and semi-implicit time advance in the NIMROD simulation code [A. H. Glasser et al., Plasma Phys. Controlled Fusion 41, A747 (1999)]. Simulation of a resistive magnetohydrodynamics mode in a shaped toroidal tokamak equilibrium demonstrates computation with disparate time scales, simulations of discharge 87009 in the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] confirm an analytic scaling for the temporal evolution of an ideal mode subject to plasma-β increasing beyond marginality, and a spherical torus simulation demonstrates nonlinear free-boundary capabilities. A comparison of numerical results on magnetic relaxation finds the n=1 mode and flux amplification in spheromaks to be very closely related to the m=1 dynamo modes and magnetic reversal in reversed-field pinch configurations. Advances in local and nonlocal closure relations developed for modeling kinetic effects in fluid simulation are also described.

Comparisons of anomalous and collisional radial transport with a continuum kinetic edge code

NASA Astrophysics Data System (ADS)

Bodi, K.; Krasheninnikov, S.; Cohen, R.; Rognlien, T.

2009-05-01

Modeling of anomalous (turbulence-driven) radial transport in controlled-fusion plasmas is necessary for long-time transport simulations. Here the focus is continuum kinetic edge codes such as the (2-D, 2-V) transport version of TEMPEST, NEO, and the code being developed by the Edge Simulation Laboratory, but the model also has wider application. Our previously developed anomalous diagonal transport matrix model with velocity-dependent convection and diffusion coefficients allows contact with typical fluid transport models (e.g., UEDGE). Results are presented that combine the anomalous transport model and collisional transport owing to ion drift orbits utilizing a Krook collision operator that conserves density and energy. Comparison is made of the relative magnitudes and possible synergistic effects of the two processes for typical tokamak device parameters.

Global two-fluid turbulence simulations of L-H transitions and edge localized mode dynamics in the COMPASS-D tokamak

NASA Astrophysics Data System (ADS)

Thyagaraja, A.; Valovič, M.; Knight, P. J.

2010-04-01

It is shown that the transition from L-mode to H-mode regimes in tokamaks can be reproduced using a two-fluid, fully electromagnetic, plasma model when a suitable particle sink is added at the edge. Such a model is implemented in the CUTIE code [A. Thyagaraja et al., Eur. J. Mech. B/Fluids 23, 475 (2004)] and is illustrated on plasma parameters that mimic those in the COMPASS-D tokamak with electron cyclotron resonance heating [Fielding et al., Plasma Phys. Contr. Fusion 42, A191 (2000)]. In particular, it is shown that holding the heating power, current, and magnetic field constant and increasing the fuelling rate to raise the plasma density leads spontaneously to the formation of an edge transport barrier (ETB) which occurs going from low to higher density experimentally. In the following quiescent period in which the stored energy of the plasma rises linearly with time, a dynamical transition occurs in the simulation with the appearance of features resembling strong edge localized modes. The simulation qualitatively reproduces many features observed in the experiment. Its relative robustness suggests that some, at least of the observed characteristics of ETBs and L-H transitions, can be captured in the global electromagnetic turbulence model.

XGC developments for a more efficient XGC-GENE code coupling

NASA Astrophysics Data System (ADS)

Dominski, Julien; Hager, Robert; Ku, Seung-Hoe; Chang, Cs

2017-10-01

In the Exascale Computing Program, the High-Fidelity Whole Device Modeling project initially aims at delivering a tightly-coupled simulation of plasma neoclassical and turbulence dynamics from the core to the edge of the tokamak. To permit such simulations, the gyrokinetic codes GENE and XGC will be coupled together. Numerical efforts are made to improve the numerical schemes agreement in the coupling region. One of the difficulties of coupling those codes together is the incompatibility of their grids. GENE is a continuum grid-based code and XGC is a Particle-In-Cell code using unstructured triangular mesh. A field-aligned filter is thus implemented in XGC. Even if XGC originally had an approximately field-following mesh, this field-aligned filter permits to have a perturbation discretization closer to the one solved in the field-aligned code GENE. Additionally, new XGC gyro-averaging matrices are implemented on a velocity grid adapted to the plasma properties, thus ensuring same accuracy from the core to the edge regions.

Three-dimensional simulations of plasma turbulence in the RFX-mod scrape-off layer and comparison with experimental measurements

NASA Astrophysics Data System (ADS)

Riva, Fabio; Vianello, Nicola; Spolaore, Monica; Ricci, Paolo; Cavazzana, Roberto; Marrelli, Lionello; Spagnolo, Silvia

2018-02-01

The tokamak scrape-off layer (SOL) plasma dynamics is investigated in a circular limiter configuration with a low edge safety factor. Focusing on the experimental parameters of two ohmic tokamak inner-wall limited plasma discharges in RFX-mod [Sonato et al., Fusion Eng. Des. 74, 97 (2005)], nonlinear SOL plasma simulations are performed with the GBS code [Ricci et al., Plasma Phys. Controlled Fusion 54, 124047 (2012)]. The numerical results are compared with the experimental measurements, assessing the reliability of the GBS model in describing the RFX-mod SOL plasma dynamics. It is found that the simulations are able to quantitatively reproduce the RFX-mod experimental measurements of the electron plasma density, electron temperature, and ion saturation current density (jsat) equilibrium profiles. Moreover, there are indications that the turbulent transport is driven by the same instability in the simulations and in the experiment, with coherent structures having similar statistical properties. On the other hand, it is found that the simulation results are not able to correctly reproduce the floating potential equilibrium profile and the jsat fluctuation level. It is likely that these discrepancies are, at least in part, related to simulating only the tokamak SOL region, without including the plasma dynamics inside the last close flux surface, and to the limits of applicability of the drift approximation. The turbulence drive is then identified from the nonlinear simulations and with the linear theory. It results that the inertial drift wave is the instability driving most of the turbulent transport in the considered discharges.

TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry

DOE PAGES

X. Q. Xu; Bodi, K.; Cohen, R. H.; ...

2010-05-28

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate themore » transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.« less

TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

X. Q. Xu; Bodi, K.; Cohen, R. H.

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate themore » transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.« less

Neoclassical orbit calculations with a full-f code for tokamak edge plasmas

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Cohen, R. H.; Dorr, M.; Hittinger, J.; Xu, X. Q.; Collela, P.; Martin, D.

2008-11-01

Ion distribution function modifications are considered for the case of neoclassical orbit widths comparable to plasma radial-gradient scale-lengths. Implementation of proper boundary conditions at divertor plates in the continuum TEMPEST code, including the effect of drifts in determining the direction of total flow, enables such calculations in single-null divertor geometry, with and without an electrostatic potential. The resultant poloidal asymmetries in densities, temperatures, and flows are discussed. For long-time simulations, a slow numerical instability develops, even in simplified (circular) geometry with no endloss, which aids identification of the mixed treatment of parallel and radial convection terms as the cause. The new Edge Simulation Laboratory code, expected to be operational, has algorithmic refinements that should address the instability. We will present any available results from the new code on this problem as well as geodesic acoustic mode tests.

Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER

DOE PAGES

Chang, Choong Seock; Ku, Seung -Hoe; Loarte, Alberto; ...

2017-07-11

Here, the XGC1 edge gyrokinetic code is used to study the width of the heat-flux to divertor plates in attached plasma condition. The flux-driven simulation is performed until an approximate power balance is achieved between the heat-flux across the steep pedestal pressure gradient and the heat-flux on the divertor plates.

Continuum kinetic modeling of the tokamak plasma edge

DOE PAGES

Dorf, M. A.; Dorr, M.; Rognlien, T.; ...

2016-03-10

In this study, the first 4D (axisymmetric) high-order continuum gyrokinetic transport simulations that span the magnetic separatrix of a tokamak are presented. The modeling is performed with the COGENT code, which is distinguished by fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasmatransport and the complex X-point divertor geometry with high accuracy. The calculations take into account the effects of fully nonlinear Fokker-Plank collisions, electrostatic potential variations, and anomalous radial transport. Topics discussed include: (a) ion orbit loss and the associated toroidal rotation and (b) edge plasma relaxation in the presence of anomalousmore » radial transport.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rasouli, C.; Abbasi Davani, F., E-mail: fabbasidavani@gmail.com

A series of experiments and numerical calculations have been done on the Damavand tokamak for accurate determination of equilibrium parameters, such as the plasma boundary position and shape. For this work, the pickup coils of the Damavand tokamak were recalibrated and after that a plasma boundary shape identification code was developed for analyzing the experimental data, such as magnetic probes and coils currents data. The plasma boundary position, shape and other parameters are determined by the plasma shape identification code. A free-boundary equilibrium code was also generated for comparison with the plasma boundary shape identification results and determination of requiredmore » fields to obtain elongated plasma in the Damavand tokamak.« less

Interaction of external n = 1 magnetic fields with the sawtooth instability in low- q RFX-mod and DIII-D tokamaks

DOE PAGES

Piron, C.; Martin, P.; Bonfiglio, D.; ...

2016-08-11

External n = 1 magnetic fields are applied in RFX-mod and DIII-D low safety factor Tokamak plasmas to investigate their interaction with the internal MHD dynamics and in particular with the sawtooth instability. In these experiments the applied magnetic fields cause a reduction of both the sawtooth amplitude and period, leading to an overall stabilizing effect on the oscillations. In RFX-mod sawteeth eventually disappear and are replaced by a stationary m = 1, n = 1 helical equilibrium without an increase in disruptivity. However toroidal rotation is significantly reduced in these plasmas, thus it is likely that the sawtooth mitigationmore » in these experiments is due to the combination of the helically deformed core and the reduced rotation. The former effect is qualitatively well reproduced by nonlinear MHD simulations performed with the PIXIE3D code. The results obtained in these RFX-mod experiments motivated similar ones in DIII-D L-mode diverted Tokamak plasmas at low q 95. These experiments succeeded in reproducing the sawtooth mitigation with the approach developed in RFX-mod. In DIII-D this effect is correlated with a clear increase of the n = 1 plasma response, that indicates an enhancement of the coupling to the marginally stable n = 1 external kink, as simulations with the linear MHD code IPEC suggest. A significant rotation braking in the plasma core is also observed in DIII-D. Finally, numerical calculations of the neoclassical toroidal viscosity (NTV) carried out with PENT identify this torque as a possible contributor for this effect.« less

Separation of Evans and Hiro currents in VDE of tokamak plasma

NASA Astrophysics Data System (ADS)

Galkin, Sergei A.; Svidzinski, V. A.; Zakharov, L. E.

2014-10-01

Progress on the Disruption Simulation Code (DSC-3D) development and benchmarking will be presented. The DSC-3D is one-fluid nonlinear time-dependent MHD code, which utilizes fully 3D toroidal geometry for the first wall, pure vacuum and plasma itself, with adaptation to the moving plasma boundary and accurate resolution of the plasma surface current. Suppression of fast magnetosonic scale by the plasma inertia neglecting will be demonstrated. Due to code adaptive nature, self-consistent plasma surface current modeling during non-linear dynamics of the Vertical Displacement Event (VDE) is accurately provided. Separation of the plasma surface current on Evans and Hiro currents during simulation of fully developed VDE, then the plasma touches in-vessel tiles, will be discussed. Work is supported by the US DOE SBIR Grant # DE-SC0004487.

Hiro and Evans currents in Vertical Disruption Event

NASA Astrophysics Data System (ADS)

Zakharov, Leonid; Xujing Li Team; Sergei Galkin Team

2014-10-01

The notion of Tokamak Magneto-Hydrodynamics (TMHD), which explicitly reflects the anisotropy of a high temperature tokamak plasma is introduced. The set of TMHD equations is formulated for simulations of macroscopic plasma dynamics and disruptions in tokamaks. Free from the Courant restriction on the time step, this set of equations is appropriate for high performance plasmas and does not require any extension of the MHD plasma model. At the same time, TMHD requires the use of magnetic field aligned numerical grids. The TMHD model was used for creation of theory of the Wall Touching Kink and Vertical Modes (WTKM and WTVM), prediction of Hiro and Evans currents, design of an innovative diagnostics for Hiro current measurements, installed on EAST device. While Hiro currents have explained the toroidal asymmetry in the plasma current measurements in JET disruptions, the Evans currents explain the tile current measurements in tokamaks. The recently developed Vertical Disruption Code (VDE) have demonstrated 5 regimes of VDE and confirmed the generation of both Hiro and Evans currents. The results challenge the 24 years long misinterpretation of the tile currents in tokamaks as ``halo'' currents, which were a product of misuse of equilibrium reconstruction for VDE. This work is supported by US DoE Contract No. DE-AC02-09-CH1146.

A methodology for the rigorous verification of plasma simulation codes

NASA Astrophysics Data System (ADS)

Riva, Fabio

2016-10-01

The methodology used to assess the reliability of numerical simulation codes constitutes the Verification and Validation (V&V) procedure. V&V is composed by two separate tasks: the verification, which is a mathematical issue targeted to assess that the physical model is correctly solved, and the validation, which determines the consistency of the code results, and therefore of the physical model, with experimental data. In the present talk we focus our attention on the verification, which in turn is composed by the code verification, targeted to assess that a physical model is correctly implemented in a simulation code, and the solution verification, that quantifies the numerical error affecting a simulation. Bridging the gap between plasma physics and other scientific domains, we introduced for the first time in our domain a rigorous methodology for the code verification, based on the method of manufactured solutions, as well as a solution verification based on the Richardson extrapolation. This methodology was applied to GBS, a three-dimensional fluid code based on a finite difference scheme, used to investigate the plasma turbulence in basic plasma physics experiments and in the tokamak scrape-off layer. Overcoming the difficulty of dealing with a numerical method intrinsically affected by statistical noise, we have now generalized the rigorous verification methodology to simulation codes based on the particle-in-cell algorithm, which are employed to solve Vlasov equation in the investigation of a number of plasma physics phenomena.

Gyrokinetic simulations and experiment

NASA Astrophysics Data System (ADS)

Ross, David W.; Bravenec, R. V.; Dorland, W.

2002-11-01

Nonlinear gyrokinetic simulations with the code GS2 have been carried out in an effort to predict transport fluxes and fluctuation levels in the tokamaks DIII-D and Alcator C-Mod.(W. Dorland et al. in Fusion Energy 2000 (International Atomic Energy Agency, Vienna, 2000).)^,( W. Ross and W. Dorland, submitted to Phys. Plasmas (2002).) These simulations account for full electron dynamics and, in some instances, electromagnetic waves.( D. W. Ross, W. Dorland, and B. N. Rogers, Bull. Am. Phys. Soc. 46, 115 (2001).) Here, some issues of the necessary resolution, precision and wave number range are examined in connection with the experimental comparisons and parameter scans.

Multi-scale gyrokinetic simulation of Alcator C-Mod tokamak discharges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howard, N. T., E-mail: nthoward@psfc.mit.edu; White, A. E.; Greenwald, M.

2014-03-15

Alcator C-Mod tokamak discharges have been studied with nonlinear gyrokinetic simulation simultaneously spanning both ion and electron spatiotemporal scales. These multi-scale simulations utilized the gyrokinetic model implemented by GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and the approximation of reduced electron mass (μ = (m{sub D}/m{sub e}){sup .5} = 20.0) to qualitatively study a pair of Alcator C-Mod discharges: a low-power discharge, previously demonstrated (using realistic mass, ion-scale simulation) to display an under-prediction of the electron heat flux and a high-power discharge displaying agreement with both ion and electron heat flux channels [N. T. Howard et al.,more » Nucl. Fusion 53, 123011 (2013)]. These multi-scale simulations demonstrate the importance of electron-scale turbulence in the core of conventional tokamak discharges and suggest it is a viable candidate for explaining the observed under-prediction of electron heat flux. In this paper, we investigate the coupling of turbulence at the ion (k{sub θ}ρ{sub s}∼O(1.0)) and electron (k{sub θ}ρ{sub e}∼O(1.0)) scales for experimental plasma conditions both exhibiting strong (high-power) and marginally stable (low-power) low-k (k{sub θ}ρ{sub s} < 1.0) turbulence. It is found that reduced mass simulation of the plasma exhibiting marginally stable low-k turbulence fails to provide even qualitative insight into the turbulence present in the realistic plasma conditions. In contrast, multi-scale simulation of the plasma condition exhibiting strong turbulence provides valuable insight into the coupling of the ion and electron scales.« less

Magnetic flux pumping mechanism prevents sawtoothing in 3D nonlinear MHD simulations of tokamak plasmas

NASA Astrophysics Data System (ADS)

Krebs, Isabel; Jardin, Stephen C.; Guenter, Sibylle; Lackner, Karl; Hoelzl, Matthias; Strumberger, Erika; Ferraro, Nate

2017-10-01

3D nonlinear MHD simulations of tokamak plasmas have been performed in toroidal geometry by means of the high-order finite element code M3D-C1. The simulations are set up such that the safety factor on axis (q0) is driven towards values below unity. As reported in and the resulting asymptotic states either exhibit sawtooth-like reconnection cycling or they are sawtooth-free. In the latter cases, a self-regulating magnetic flux pumping mechanism, mainly provided by a saturated quasi-interchange instability via a dynamo effect, redistributes the central current density so that the central safety factor profile is flat and q0 1 . Sawtoothing is prevented if β is sufficiently high to allow for the necessary amount of flux pumping to counterbalance the tendency of the current density profile to centrally peak. We present the results of 3D nonlinear simulations based on specific types of experimental discharges and analyze their asymptotic behavior. A set of cases is presented where aspects of the current ramp-up phase of Hybrid ASDEX Upgrade discharges are mimicked. Another set of simulations is based on low-qedge discharges in DIII-D.

Simulating the effects of stellarator geometry on gyrokinetic drift-wave turbulence

NASA Astrophysics Data System (ADS)

Baumgaertel, Jessica Ann

Nuclear fusion is a clean, safe form of energy with abundant fuel. In magnetic fusion energy (MFE) experiments, the plasma fuel is confined by magnetic fields at very high temperatures and densities. One fusion reactor design is the non-axisymmetric, torus-shaped stellarator. Its fully-3D fields have advantages over the simpler, better-understood axisymmetric tokamak, including the ability to optimize magnetic configurations for desired properties, such as lower transport (longer confinement time). Turbulence in the plasma can break MFE confinement. While turbulent transport is known to cause a significant amount of heat loss in tokamaks, it is a new area of research in stellarators. Gyrokinetics is a good mathematical model of the drift-wave instabilities that cause turbulence. Multiple gyrokinetic turbulence codes that had great success comparing to tokamak experiments are being converted for use with stellarator geometry. This thesis describes such adaptations of the gyrokinetic turbulence code, GS2. Herein a new computational grid generator and upgrades to GS2 itself are described, tested, and benchmarked against three other gyrokinetic codes. Using GS2, detailed linear studies using the National Compact Stellarator Experiment (NCSX) geometry were conducted. The first compares stability in two equilibria with different β=(plasma pressure)/(magnetic pressure). Overall, the higher β case was more stable than the lower β case. As high β is important for MFE experiments, this is encouraging. The second compares NCSX linear stability to a tokamak case. NCSX was more stable with a 20% higher critical temperature gradient normalized by the minor radius, suggesting that the fusion power might be enhanced by ˜ 50%. In addition, the first nonlinear, non-axisymmetric GS2 simulations are presented. Finally, linear stability of two locations in a W7-AS plasma were compared. The experimentally-measured parameters used were from a W7-AS shot in which measured heat fluxes match neoclassical theory predictions at inner radii, but are too large for neoclassical predictions at outer radii. Results from GS2 linear simulations show that the outer location has higher gyrokinetic instability growth rates than at the inner one. Mixing-length estimates of the heat flux are within a factor of 3 of the experimental measurements, indicating that gyrokinetic turbulence may be responsible for the higher transport measured by the experiment in the outer regions. Future nonlinear simulations can explore this question in more detail. This work is supported by the Princeton Plasma Physics Laboratory, which is operated by Princeton University for the U.S. Department of Energy under Contract No. DE-AC02-09CH11466, and the SciDAC Center for the Study of Plasma Microturbulence.

A collision scheme for hybrid fluid-particle simulation of plasmas

NASA Astrophysics Data System (ADS)

Nguyen, Christine; Lim, Chul-Hyun; Verboncoeur, John

2006-10-01

Desorption phenomena at the wall of a tokamak can lead to the introduction of impurities at the edge of a thermonuclear plasma. In particular, the use of carbon as a constituent of the tokamak wall, as planned for ITER, requires the study of carbon and hydrocarbon transport in the plasma, including understanding of collisional interaction with the plasma. These collisions can result in new hydrocarbons, hydrogen, secondary electrons and so on. Computational modeling is a primary tool for studying these phenomena. XOOPIC [1] and OOPD1 are widely used computer modeling tools for the simulation of plasmas. Both are particle type codes. Particle simulation gives more kinetic information than fluid simulation, but more computation time is required. In order to reduce this disadvantage, hybrid simulation has been developed, and applied to the modeling of collisions. Present particle simulation tools such as XOOPIC and OODP1 employ a Monte Carlo model for the collisions between particle species and a neutral background gas defined by its temperature and pressure. In fluid-particle hybrid plasma models, collisions include combinations of particle and fluid interactions categorized by projectile-target pairing: particle-particle, particle-fluid, and fluid-fluid. For verification of this hybrid collision scheme, we compare simulation results to analytic solutions for classical plasma models. [1] Verboncoeur et al. Comput. Phys. Comm. 87, 199 (1995).

Advanced simulation of mixed-material erosion/evolution and application to low and high-Z containing plasma facing components

NASA Astrophysics Data System (ADS)

Brooks, J. N.; Hassanein, A.; Sizyuk, T.

2013-07-01

Plasma interactions with mixed-material surfaces are being analyzed using advanced modeling of time-dependent surface evolution/erosion. Simulations use the REDEP/WBC erosion/redeposition code package coupled to the HEIGHTS package ITMC-DYN mixed-material formation/response code, with plasma parameter input from codes and data. We report here on analysis for a DIII-D Mo/C containing tokamak divertor. A DIII-D/DiMES probe experiment simulation predicts that sputtered molybdenum from a 1 cm diameter central spot quickly saturates (˜4 s) in the 5 cm diameter surrounding carbon probe surface, with subsequent re-sputtering and transport to off-probe divertor regions, and with high (˜50%) redeposition on the Mo spot. Predicted Mo content in the carbon agrees well with post-exposure probe data. We discuss implications and mixed-material analysis issues for Be/W mixing at the ITER outer divertor, and Li, C, Mo mixing at an NSTX divertor.

Tempest Neoclassical Simulation of Fusion Edge Plasmas

NASA Astrophysics Data System (ADS)

Xu, X. Q.; Xiong, Z.; Cohen, B. I.; Cohen, R. H.; Dorr, M.; Hittinger, J.; Kerbel, G. D.; Nevins, W. M.; Rognlien, T. D.

2006-04-01

We are developing a continuum gyrokinetic full-F code, TEMPEST, to simulate edge plasmas. The geometry is that of a fully diverted tokamak and so includes boundary conditions for both closed magnetic flux surfaces and open field lines. The code, presently 4-dimensional (2D2V), includes kinetic ions and electrons, a gyrokinetic Poisson solver for electric field, and the nonlinear Fokker-Planck collision operator. Here we present the simulation results of neoclassical transport with Boltzmann electrons. In a large aspect ratio circular geometry, excellent agreement is found for neoclassical equilibrium with parallel flows in the banana regime without a temperature gradient. In divertor geometry, it is found that the endloss of particles and energy induces pedestal-like density and temperature profiles inside the magnetic separatrix and parallel flow stronger than the neoclassical predictions in the SOL. The impact of the X-point divertor geometry on the self-consistent electric field and geo-acoustic oscillations will be reported. We will also discuss the status of extending TEMPEST into a 5-D code.

The build-up of energetic electrons triggering electron cyclotron emission bursts due to a magnetohydrodynamic mode at the edge of tokamaks

DOE PAGES

Li, Erzhong; Austin, Max E.; White, R. B.; ...

2017-08-21

Intense bursts of electron cyclotron emission (ECE) triggered by magnetohydrodynamic (MHD) instabilities such as edge localized modes (ELMs) have been observed on many tokamaks. On the DIII-D tokamak, it is found that an MHD mode is observed to trigger the ECE bursts in the low collisionality regime at the plasma edge. ORBIT-code simulations have shown that energetic electrons build up due to an interaction between barely trapped electrons with an MHD mode (f = 50 kHz for current case). The energetic tail of the electron distribution function develops a bump within several microseconds for this collisionless case. This behavior dependsmore » on the competition between the perturbing MHD mode and slowing down and pitch angle scattering due to collisions. As a result, for typical DIII-D parameters, the calculated ECE radiation transport predicted by ORBIT is in excellent agreement with ECE measurements, clarifying the electron dynamics of the ECE bursts for the first time.« less

Simulation of the ELMs triggering by lithium pellet on EAST tokamak using BOUT + +

NASA Astrophysics Data System (ADS)

Wang, Y. M.; Xu, X. Q.; Wang, Z.; Sun, Z.; Hu, J. S.; Gao, X.

2017-10-01

A new lithium granule injector (LGI) was developed on EAST. Using the LGI, lithium granules can be efficiently injected into EAST tokamak with the granule radius 0.2-1 mm and the granules velocity 30-110 m/s. ELM pacing was realized during EAST shot #70123 at time window from 4.4-4.7s, the average velocity of the pellet was 75 m/s and the average injection rate is at 99Hz. The BOUT + + 6-field electromagnetic turbulence code has been used to simulate the ELM pacing process. A neutral gas shielding (NGS) model has been implemented during the pellet ablation process. The neutral transport code is used to evaluate the ionized electron and Li ion densities with the charge exchange as a dominant factor in the neutral cloud diffusion process. The snapshot plasma profiles during the pellet ablation and toroidal symmetrization process are used in the 6-field turbulence code to evaluate the impact of the pellets on ELMs. Destabilizing effects of the peeling-ballooning modes are found with lithium pellet injection, which is consistent with the experimental results. A scan of the pellet size, shape and the injection velocity will be conducted, which will benefit the pellet injection design in both the present and future devices. Prepared by LLNL under Contract DE-AC52-07NA27344 and this work is supported by the National Natural Science Fonudation of China (Grant No. 11505221) and China Scholarship Council (Grant No. 201504910132).

SciDAC GSEP: Gyrokinetic Simulation of Energetic Particle Turbulence and Transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lin, Zhihong

Energetic particle (EP) confinement is a key physics issue for burning plasma experiment ITER, the crucial next step in the quest for clean and abundant energy, since ignition relies on self-heating by energetic fusion products (α-particles). Due to the strong coupling of EP with burning thermal plasmas, plasma confinement property in the ignition regime is one of the most uncertain factors when extrapolating from existing fusion devices to the ITER tokamak. EP population in current tokamaks are mostly produced by auxiliary heating such as neutral beam injection (NBI) and radio frequency (RF) heating. Remarkable progress in developing comprehensive EP simulationmore » codes and understanding basic EP physics has been made by two concurrent SciDAC EP projects GSEP funded by the Department of Energy (DOE) Office of Fusion Energy Science (OFES), which have successfully established gyrokinetic turbulence simulation as a necessary paradigm shift for studying the EP confinement in burning plasmas. Verification and validation have rapidly advanced through close collaborations between simulation, theory, and experiment. Furthermore, productive collaborations with computational scientists have enabled EP simulation codes to effectively utilize current petascale computers and emerging exascale computers. We review here key physics progress in the GSEP projects regarding verification and validation of gyrokinetic simulations, nonlinear EP physics, EP coupling with thermal plasmas, and reduced EP transport models. Advances in high performance computing through collaborations with computational scientists that enable these large scale electromagnetic simulations are also highlighted. These results have been widely disseminated in numerous peer-reviewed publications including many Phys. Rev. Lett. papers and many invited presentations at prominent fusion conferences such as the biennial International Atomic Energy Agency (IAEA) Fusion Energy Conference and the annual meeting of the American Physics Society, Division of Plasma Physics (APS-DPP).« less

M3D-K Simulations of Beam-Driven Alfven Eigenmodes in ASDEX-U

NASA Astrophysics Data System (ADS)

Wang, Ge; Fu, Guoyong; Lauber, Philipp; Schneller, Mirjam

2013-10-01

Core-localized Alfven eigenmodes are often observed in neutral beam-heated plasma in ASDEX-U tokamak. In this work, hybrid simulations with the global kinetic/MHD hybrid code M3D-K have been carried out to investigate the linear stability and nonlinear dynamics of beam-driven Alfven eigenmodes using experimental parameters and profiles of an ASDEX-U discharge. The safety factor q profile is weakly reversed with minimum q value about qmin = 3.0. The simulation results show that the n = 3 mode transits from a reversed shear Alfven eigenmode (RSAE) to a core-localized toroidal Alfven eigenmode (TAE) as qmin drops from 3.0 to 2.79, consistent with results from the stability code NOVA as well as the experimental measurement. The M3D-K results are being compared with those of the linear gyrokinetic stability code LIGKA for benchmark. The simulation results will also be compared with the measured mode frequency and mode structure. This work was funded by the Max-Planck/Princeton Center for Plasma Physics.

Neoclassical tearing mode seeding by coupling with infernal modes in low-shear tokamaks

NASA Astrophysics Data System (ADS)

Kleiner, A.; Graves, J. P.; Brunetti, D.; Cooper, W. A.; Halpern, F. D.; Luciani, J.-F.; Lütjens, H.

2016-09-01

A numerical and an analytical study of the triggering of resistive MHD modes in tokamak plasmas with low magnetic shear core is presented. Flat q profiles give rise to fast growing pressure driven MHD modes, such as infernal modes. It has been shown that infernal modes drive fast growing islands on neighbouring rational surfaces. Numerical simulations of such instabilities in a MAST-like configuration are performed with the initial value stability code XTOR-2F in the resistive frame. The evolution of magnetic islands are computed from XTOR-2F simulations and an analytical model is developed based on Rutherford’s theory in combination with a model of resistive infernal modes. The parameter {{Δ }\\prime} is extended from the linear phase to the non-linear phase. Additionally, the destabilising contribution due to a helically perturbed bootstrap current is considered. Comparing the numerical XTOR-2F simulations to the model, we find that coupling has a strong destabilising effect on (neoclassical) tearing modes and is able to seed 2/1 magnetic islands in situations when the standard NTM theory predicts stability.

Progress with the COGENT Edge Kinetic Code: Implementing the Fokker-Plank Collision Operator

DOE PAGES

Dorf, M. A.; Cohen, R. H.; Dorr, M.; ...

2014-06-20

Here, COGENT is a continuum gyrokinetic code for edge plasma simulations being developed by the Edge Simulation Laboratory collaboration. The code is distinguished by application of a fourth-order finite-volume (conservative) discretization, and mapped multiblock grid technology to handle the geometric complexity of the tokamak edge. The distribution function F is discretized in v∥ – μ (parallel velocity – magnetic moment) velocity coordinates, and the code presently solves an axisymmetric full-f gyro-kinetic equation coupled to the long-wavelength limit of the gyro-Poisson equation. COGENT capabilities are extended by implementing the fully nonlinear Fokker-Plank operator to model Coulomb collisions in magnetized edge plasmas.more » The corresponding Rosenbluth potentials are computed by making use of a finite-difference scheme and multipole-expansion boundary conditions. Details of the numerical algorithms and results of the initial verification studies are discussed. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)« less

Interaction of external n  =  1 magnetic fields with the sawtooth instability in low-q RFX-mod and DIII-D tokamaks

NASA Astrophysics Data System (ADS)

Piron, C.; Martin, P.; Bonfiglio, D.; Hanson, J.; Logan, N. C.; Paz-Soldan, C.; Piovesan, P.; Turco, F.; Bialek, J.; Franz, P.; Jackson, G.; Lanctot, M. J.; Navratil, G. A.; Okabayashi, M.; Strait, E.; Terranova, D.; Turnbull, A.

2016-10-01

External n  =  1 magnetic fields are applied in RFX-mod and DIII-D low safety factor Tokamak plasmas to investigate their interaction with the internal MHD dynamics and in particular with the sawtooth instability. In these experiments the applied magnetic fields cause a reduction of both the sawtooth amplitude and period, leading to an overall stabilizing effect on the oscillations. In RFX-mod sawteeth eventually disappear and are replaced by a stationary m  =  1, n  =  1 helical equilibrium without an increase in disruptivity. However toroidal rotation is significantly reduced in these plasmas, thus it is likely that the sawtooth mitigation in these experiments is due to the combination of the helically deformed core and the reduced rotation. The former effect is qualitatively well reproduced by nonlinear MHD simulations performed with the PIXIE3D code. The results obtained in these RFX-mod experiments motivated similar ones in DIII-D L-mode diverted Tokamak plasmas at low q 95. These experiments succeeded in reproducing the sawtooth mitigation with the approach developed in RFX-mod. In DIII-D this effect is correlated with a clear increase of the n  =  1 plasma response, that indicates an enhancement of the coupling to the marginally stable n  =  1 external kink, as simulations with the linear MHD code IPEC suggest. A significant rotation braking in the plasma core is also observed in DIII-D. Numerical calculations of the neoclassical toroidal viscosity (NTV) carried out with PENT identify this torque as a possible contributor for this effect.

Energy balance in TM-1-MH Tokamak (ohmical heating)

NASA Astrophysics Data System (ADS)

Stoeckel, J.; Koerbel, S.; Kryska, L.; Kopecky, V.; Dadalec, V.; Datlov, J.; Jakubka, K.; Magula, P.; Zacek, F.; Pereverzev, G. V.

1981-10-01

Plasma in the TM-1-MH Tokamak was experimentally studied in the parameter range: tor. mg. field B = 1,3 T, plasma current I sub p = 14 kA, electron density N sub E 3.10 to the 19th power cubic meters. The two numerical codes are available for the comparison with experimental data. TOKATA-code solves simplified energy balance equations for electron and ion components. TOKSAS-code solves the detailed energy balance of the ion component.

Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment

NASA Astrophysics Data System (ADS)

Boyle, D. P.; Bell, R. E.; Kaita, R.; Lucia, M.; Schmitt, J. C.; Scotti, F.; Kubota, S.; Hansen, C.; Biewer, T. M.; Gray, T. K.

2016-10-01

The Lithium Tokamak Experiment (LTX) is a modest-sized spherical tokamak with all-metal plasma facing components (PFCs), uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma. This work presents measurements of core plasma impurity concentrations and transport in LTX. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with 2 - 4 % Li, 0.6 - 2 % C, 0.4 - 0.7 % O, and Zeff < 1.2 . Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, and neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two. However, time-independent simulations with MIST indicated that neoclassical theory did not fully capture the impurity transport and anomalous transport likely played a significant role in determining impurity profiles. Progress on additional analysis, including time-dependent impurity transport simulations and impurity measurements with liquid lithium coatings, and plans for diagnostic upgrades and future experiments in LTX- β will also be presented. This work supported by US DOE contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.

Effects of MHD instabilities on neutral beam current drive

NASA Astrophysics Data System (ADS)

Podestà, M.; Gorelenkova, M.; Darrow, D. S.; Fredrickson, E. D.; Gerhardt, S. P.; White, R. B.

2015-05-01

Neutral beam injection (NBI) is one of the primary tools foreseen for heating, current drive (CD) and q-profile control in future fusion reactors such as ITER and a Fusion Nuclear Science Facility. However, fast ions from NBI may also provide the drive for energetic particle-driven instabilities (e.g. Alfvénic modes (AEs)), which in turn redistribute fast ions in both space and energy, thus hampering the control capabilities and overall efficiency of NB-driven current. Based on experiments on the NSTX tokamak (M. Ono et al 2000 Nucl. Fusion 40 557), the effects of AEs and other low-frequency magneto-hydrodynamic instabilities on NB-CD efficiency are investigated. A new fast ion transport model, which accounts for particle transport in phase space as required for resonant AE perturbations, is utilized to obtain consistent simulations of NB-CD through the tokamak transport code TRANSP. It is found that instabilities do indeed reduce the NB-driven current density over most of the plasma radius by up to ∼50%. Moreover, the details of the current profile evolution are sensitive to the specific model used to mimic the interaction between NB ions and instabilities. Implications for fast ion transport modeling in integrated tokamak simulations are briefly discussed.

Phase space effects on fast ion distribution function modeling in tokamaks

NASA Astrophysics Data System (ADS)

Podestà, M.; Gorelenkova, M.; Fredrickson, E. D.; Gorelenkov, N. N.; White, R. B.

2016-05-01

Integrated simulations of tokamak discharges typically rely on classical physics to model energetic particle (EP) dynamics. However, there are numerous cases in which energetic particles can suffer additional transport that is not classical in nature. Examples include transport by applied 3D magnetic perturbations and, more notably, by plasma instabilities. Focusing on the effects of instabilities, ad-hoc models can empirically reproduce increased transport, but the choice of transport coefficients is usually somehow arbitrary. New approaches based on physics-based reduced models are being developed to address those issues in a simplified way, while retaining a more correct treatment of resonant wave-particle interactions. The kick model implemented in the tokamak transport code TRANSP is an example of such reduced models. It includes modifications of the EP distribution by instabilities in real and velocity space, retaining correlations between transport in energy and space typical of resonant EP transport. The relevance of EP phase space modifications by instabilities is first discussed in terms of predicted fast ion distribution. Results are compared with those from a simple, ad-hoc diffusive model. It is then shown that the phase-space resolved model can also provide additional insight into important issues such as internal consistency of the simulations and mode stability through the analysis of the power exchanged between energetic particles and the instabilities.

Phase space effects on fast ion distribution function modeling in tokamaks

DOE Data Explorer

White, R. B. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Podesta, M. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Gorelenkova, M. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Fredrickson, E. D. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Gorelenkov, N. N. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

2016-06-01

Integrated simulations of tokamak discharges typically rely on classical physics to model energetic particle (EP) dynamics. However, there are numerous cases in which energetic particles can suffer additional transport that is not classical in nature. Examples include transport by applied 3D magnetic perturbations and, more notably, by plasma instabilities. Focusing on the effects of instabilities, ad-hoc models can empirically reproduce increased transport, but the choice of transport coefficients is usually somehow arbitrary. New approaches based on physics-based reduced models are being developed to address those issues in a simplified way, while retaining a more correct treatment of resonant wave-particle interactions. The kick model implemented in the tokamak transport code TRANSP is an example of such reduced models. It includes modifications of the EP distribution by instabilities in real and velocity space, retaining correlations between transport in energy and space typical of resonant EP transport. The relevance of EP phase space modifications by instabilities is first discussed in terms of predicted fast ion distribution. Results are compared with those from a simple, ad-hoc diffusive model. It is then shown that the phase-space resolved model can also provide additional insight into important issues such as internal consistency of the simulations and mode stability through the analysis of the power exchanged between energetic particles and the instabilities.

Effects of MHD instabilities on neutral beam current drive

DOE PAGES

Podestà, M.; Gorelenkova, M.; Darrow, D. S.; ...

2015-04-17

One of the primary tools foreseen for heating, current drive (CD) and q-profile control in future fusion reactors such as ITER and a Fusion Nuclear Science Facility is the neutral beam injection (NBI). However, fast ions from NBI may also provide the drive for energetic particle-driven instabilities (e.g. Alfvénic modes (AEs)), which in turn redistribute fast ions in both space and energy, thus hampering the control capabilities and overall efficiency of NB-driven current. Based on experiments on the NSTX tokamak (M. Ono et al 2000 Nucl. Fusion 40 557), the effects of AEs and other low-frequency magneto-hydrodynamic instabilities on NB-CDmore » efficiency are investigated. When looking at the new fast ion transport model, which accounts for particle transport in phase space as required for resonant AE perturbations, is utilized to obtain consistent simulations of NB-CD through the tokamak transport code TRANSP. It is found that instabilities do indeed reduce the NB-driven current density over most of the plasma radius by up to ~50%. Moreover, the details of the current profile evolution are sensitive to the specific model used to mimic the interaction between NB ions and instabilities. Finally, implications for fast ion transport modeling in integrated tokamak simulations are briefly discussed.« less

Effects of Equilibrium Toroidal Flow on Locked Mode and Plasma Response in a Tokamak

NASA Astrophysics Data System (ADS)

Zhu, Ping; Huang, Wenlong; Yan, Xingting

2016-10-01

It is widely believed that plasma flow plays significant roles in regulating the processes of mode locking and plasma response in a tokamak in presence of external resonant magnetic perturbations (RMPs). Recently a common analytic relation for both locked mode and plasma response has been developed based on the steady-state solution to the coupled dynamic system of magnetic island evolution and torque balance. The analytic relation predicts the size of the magnetic island of a locked mode or a static nonlinear plasma response for a given RMP amplitude, and rigorously proves a screening effect of the equilibrium toroidal flow. To test the theory, we solve for the locked mode and the nonlinear plasma response in presence of RMP for a circular-shaped limiter tokamak equilibrium with constant toroidal flow, using the initial-value, full MHD simulation code NIMROD. The comparison between the simulation results and the theory prediction, in terms of the quantitative screening effects of equilibrium toroidal flow, will be reported and discussed. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB101004, the 100 Talent Program of the Chinese Academy of Sciences, and U.S. Department of Energy Grants DE-FG02-86ER53218 and DE-FC02-08ER54975.

Comparing nonlinear MHD simulations of low-aspect-ratio RFPs to RELAX experiments

NASA Astrophysics Data System (ADS)

McCollam, K. J.; den Hartog, D. J.; Jacobson, C. M.; Sovinec, C. R.; Masamune, S.; Sanpei, A.

2016-10-01

Standard reversed-field pinch (RFP) plasmas provide a nonlinear dynamical system as a validation domain for numerical MHD simulation codes, with applications in general toroidal confinement scenarios including tokamaks. Using the NIMROD code, we simulate the nonlinear evolution of RFP plasmas similar to those in the RELAX experiment. The experiment's modest Lundquist numbers S (as low as a few times 104) make closely matching MHD simulations tractable given present computing resources. Its low aspect ratio ( 2) motivates a comparison study using cylindrical and toroidal geometries in NIMROD. We present initial results from nonlinear single-fluid runs at S =104 for both geometries and a range of equilibrium parameters, which preliminarily show that the magnetic fluctuations are roughly similar between the two geometries and between simulation and experiment, though there appear to be some qualitative differences in their temporal evolution. Runs at higher S are planned. This work is supported by the U.S. DOE and by the Japan Society for the Promotion of Science.

Microturbulence studies of pulsed poloidal current drive discharges in the reversed field pinch

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmody, D., E-mail: dcarmody@wisc.edu; Pueschel, M. J.; Anderson, J. K.

2015-01-15

Experimental discharges with pulsed poloidal current drive (PPCD) in the Madison Symmetric Torus reversed field pinch are investigated using a semi-analytic equilibrium model in the gyrokinetic turbulence code GENE. PPCD cases, with plasma currents of 500 kA and 200 kA, exhibit a density-gradient-driven trapped electron mode (TEM) and an ion temperature gradient mode, respectively. Relative to expectations of tokamak core plasmas, the critical gradients for the onset of these instabilities are found to be greater by roughly a factor of the aspect ratio. A significant upshift in the nonlinear TEM transport threshold, previously found for tokamaks, is confirmed in nonlinear reversed fieldmore » pinch simulations and is roughly three times the threshold for linear instability. The simulated heat fluxes can be brought in agreement with measured diffusivities by introducing a small, resonant magnetic perturbation, thus modeling the residual fluctuations from tearing modes. These fluctuations significantly enhance transport.« less

Overview of Edge Simulation Laboratory (ESL)

NASA Astrophysics Data System (ADS)

Cohen, R. H.; Dorr, M.; Hittinger, J.; Rognlien, T.; Umansky, M.; Xiong, A.; Xu, X.; Belli, E.; Candy, J.; Snyder, P.; Colella, P.; Martin, D.; Sternberg, T.; van Straalen, B.; Bodi, K.; Krasheninnikov, S.

2006-10-01

The ESL is a new collaboration to build a full-f electromagnetic gyrokinetic code for tokamak edge plasmas using continuum methods. Target applications are edge turbulence and transport (neoclassical and anomalous), and edge-localized modes. Initially the project has three major threads: (i) verification and validation of TEMPEST, the project's initial (electrostatic) edge code which can be run in 4D (neoclassical and transport-timescale applications) or 5D (turbulence); (ii) design of the next generation code, which will include more complete physics (electromagnetics, fluid equation option, improved collisions) and advanced numerics (fully conservative, high-order discretization, mapped multiblock grids, adaptivity), and (iii) rapid-prototype codes to explore the issues attached to solving fully nonlinear gyrokinetics with steep radial gradiens. We present a brief summary of the status of each of these activities.

Advances in simulation of wave interactions with extended MHD phenomena

NASA Astrophysics Data System (ADS)

Batchelor, D.; Abla, G.; D'Azevedo, E.; Bateman, G.; Bernholdt, D. E.; Berry, L.; Bonoli, P.; Bramley, R.; Breslau, J.; Chance, M.; Chen, J.; Choi, M.; Elwasif, W.; Foley, S.; Fu, G.; Harvey, R.; Jaeger, E.; Jardin, S.; Jenkins, T.; Keyes, D.; Klasky, S.; Kruger, S.; Ku, L.; Lynch, V.; McCune, D.; Ramos, J.; Schissel, D.; Schnack, D.; Wright, J.

2009-07-01

The Integrated Plasma Simulator (IPS) provides a framework within which some of the most advanced, massively-parallel fusion modeling codes can be interoperated to provide a detailed picture of the multi-physics processes involved in fusion experiments. The presentation will cover four topics: 1) recent improvements to the IPS, 2) application of the IPS for very high resolution simulations of ITER scenarios, 3) studies of resistive and ideal MHD stability in tokamk discharges using IPS facilities, and 4) the application of RF power in the electron cyclotron range of frequencies to control slowly growing MHD modes in tokamaks and initial evaluations of optimized location for RF power deposition.

Global simulation of edge pedestal micro-instabilities

NASA Astrophysics Data System (ADS)

Wan, Weigang; Parker, Scott; Chen, Yang

2011-10-01

We study micro turbulence of the tokamak edge pedestal with global gyrokinetic particle simulations. The simulation code GEM is an electromagnetic δf code. Two sets of DIII-D experimental profiles, shot #131997 and shot #136051 are used. The dominant instabilities appear to be two kinds of modes both propagating in the electron diamagnetic direction, with comparable linear growth rates. The low n mode is at the Alfven frequency range and driven by density and ion temperature gradients. The high n mode is driven by electron temperature gradient and has a low real frequency. A β scan shows that the low n mode is electromagnetic. Frequency analysis shows that the high n mode is sometimes mixed with an ion instability. Experimental radial electric field is applied and its effects studied. We will also show some preliminary nonlinear results. We thank R. Groebner, P. Snyder and Y. Zheng for providing experimental profiles and helpful discussions.

Physics of Tokamak Plasma Start-up

NASA Astrophysics Data System (ADS)

Mueller, Dennis

2012-10-01

This tutorial describes and reviews the state-of-art in tokamak plasma start-up and its importance to next step devices such as ITER, a Fusion Nuclear Science Facility and a Tokamak/ST demo. Tokamak plasma start-up includes breakdown of the initial gas, ramp-up of the plasma current to its final value and the control of plasma parameters during those phases. Tokamaks rely on an inductive component, typically a central solenoid, which has enabled attainment of high performance levels that has enabled the construction of the ITER device. Optimizing the inductive start-up phase continues to be an area of active research, especially in regards to achieving ITER scenarios. A new generation of superconducting tokamaks, EAST and KSTAR, experiments on DIII-D and operation with JET's ITER-like wall are contributing towards this effort. Inductive start-up relies on transformer action to generate a toroidal loop voltage and successful start-up is determined by gas breakdown, avalanche physics and plasma-wall interaction. The goal of achieving steady-sate tokamak operation has motivated interest in other methods for start-up that do not rely on the central solenoid. These include Coaxial Helicity Injection, outer poloidal field coil start-up, and point source helicity injection, which have achieved 200, 150 and 100 kA respectively of toroidal current on closed flux surfaces. Other methods including merging reconnection startup and Electron Bernstein Wave (EBW) plasma start-up are being studied on various devices. EBW start-up generates a directed electron channel due to wave particle interaction physics while the other methods mentioned rely on magnetic helicity injection and magnetic reconnection which are being modeled and understood using NIMROD code simulations.

Implementation of the full viscoresistive magnetohydrodynamic equations in a nonlinear finite element code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haverkort, J.W.; Dutch Institute for Fundamental Energy Research, P.O. Box 6336, 5600 HH Eindhoven; Blank, H.J. de

Numerical simulations form an indispensable tool to understand the behavior of a hot plasma that is created inside a tokamak for providing nuclear fusion energy. Various aspects of tokamak plasmas have been successfully studied through the reduced magnetohydrodynamic (MHD) model. The need for more complete modeling through the full MHD equations is addressed here. Our computational method is presented along with measures against possible problems regarding pollution, stability, and regularity. The problem of ensuring continuity of solutions in the center of a polar grid is addressed in the context of a finite element discretization of the full MHD equations. Amore » rigorous and generally applicable solution is proposed here. Useful analytical test cases are devised to verify the correct implementation of the momentum and induction equation, the hyperdiffusive terms, and the accuracy with which highly anisotropic diffusion can be simulated. A striking observation is that highly anisotropic diffusion can be treated with the same order of accuracy as isotropic diffusion, even on non-aligned grids, as long as these grids are generated with sufficient care. This property is shown to be associated with our use of a magnetic vector potential to describe the magnetic field. Several well-known instabilities are simulated to demonstrate the capabilities of the new method. The linear growth rate of an internal kink mode and a tearing mode are benchmarked against the results of a linear MHD code. The evolution of a tearing mode and the resulting magnetic islands are simulated well into the nonlinear regime. The results are compared with predictions from the reduced MHD model. Finally, a simulation of a ballooning mode illustrates the possibility to use our method as an ideal MHD method without the need to add any physical dissipation.« less

Toward validation of a 3-D plasma turbulence model using LAPD data

NASA Astrophysics Data System (ADS)

Umansky, M. V.

2010-11-01

Detailed results from a 3-D fluid simulation of plasma turbulence are compared with experimental data from the Large Plasma Device (LAPD) at UCLA. LAPD is a magnetized plasma column experiment with a high repetition rate, allowing detailed time-and-space resolved probe data on plasma turbulence and transport. The large amount of data allows a thorough comparison with the simulation results. For the observed drift-type modes, LAPD plasmas are strongly collisional (φ*/νei1 and λei/L1), providing justification for a fluid treatment. Accordingly, the model is based on reduced Braginskii equations and is implemented in the framework of the BOUT code, originally developed at LLNL for tokamak edge plasmas. Analysis of linear plasma instabilities shows that resistive drift modes, rotation-driven interchange modes, and Kelvin-Helmholtz modes can all be important in LAPD and have comparable frequencies and growth rates. In nonlinear simulations using measured LAPD density profiles, evolution of instabilities and self-generated zonal flows results in a saturated turbulent state. Comparisons of these simulations with measurements in LAPD plasmas reveal good agreement, in particular in the frequency spectrum, spatial correlation, and amplitude probability distribution function of density fluctuations. Also, consistent with the experiment, the simulations indicate a great deal of similarity between plasma turbulence in LAPD and some features of tokamak edge turbulence. Similar to tokamak edge plasmas, density transport appears to be predominantly carried by large particle-flux events. Despite the intermittent character of the calculated turbulence, as indicated by fluctuation statistics, the turbulent particle flux is consistent with a diffusive model with diffusion coefficient close to the Bohm value.

User's manual for COAST 4: a code for costing and sizing tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sink, D. A.; Iwinski, E. M.

1979-09-01

The purpose of this report is to document the computer program COAST 4 for the user/analyst. COAST, COst And Size Tokamak reactors, provides complete and self-consistent size models for the engineering features of D-T burning tokamak reactors and associated facilities involving a continuum of performance including highly beam driven through ignited plasma devices. TNS (The Next Step) devices with no tritium breeding or electrical power production are handled as well as power producing and fissile producing fusion-fission hybrid reactors. The code has been normalized with a TFTR calculation which is consistent with cost, size, and performance data published in themore » conceptual design report for that device. Information on code development, computer implementation and detailed user instructions are included in the text.« less

Flux-driven turbulence GDB simulations of the IWL Alcator C-Mod L-mode edge compared with experiment

NASA Astrophysics Data System (ADS)

Francisquez, Manaure; Zhu, Ben; Rogers, Barrett

2017-10-01

Prior to predicting confinement regime transitions in tokamaks one may need an accurate description of L-mode profiles and turbulence properties. These features determine the heat-flux width upon which wall integrity depends, a topic of major interest for research aid to ITER. To this end our work uses the GDB model to simulate the Alcator C-Mod edge and contributes support for its use in studying critical edge phenomena in current and future tokamaks. We carried out 3D electromagnetic flux-driven two-fluid turbulence simulations of inner wall limited (IWL) C-Mod shots spanning closed and open flux surfaces. These simulations are compared with gas puff imaging (GPI) and mirror Langmuir probe (MLP) data, examining global features and statistical properties of turbulent dynamics. GDB reproduces important qualitative aspects of the C-Mod edge regarding global density and temperature profiles, within reasonable margins, and though the turbulence statistics of the simulated turbulence follow similar quantitative trends questions remain about the code's difficulty in exactly predicting quantities like the autocorrelation time A proposed breakpoint in the near SOL pressure and the posited separation between drift and ballooning dynamics it represents are examined This work was supported by DOE-SC-0010508. This research used resources of the National Energy Research Scientific Computing Center (NERSC).

Nonlinear Diamagnetic Stabilization of Double Tearing Modes in Cylindrical MHD Simulations

NASA Astrophysics Data System (ADS)

Abbott, Stephen; Germaschewski, Kai

2014-10-01

Double tearing modes (DTMs) may occur in reversed-shear tokamak configurations if two nearby rational surfaces couple and begin reconnecting. During the DTM's nonlinear evolution it can enter an ``explosive'' growth phase leading to complete reconnection, making it a possible driver for off-axis sawtooth crashes. Motivated by similarities between this behavior and that of the m = 1 kink-tearing mode in conventional tokamaks we investigate diamagnetic drifts as a possible DTM stabilization mechanism. We extend our previous linear studies of an m = 2 , n = 1 DTM in cylindrical geometry to the fully nonlinear regime using the MHD code MRC-3D. A pressure gradient similar to observed ITB profiles is used, together with Hall physics, to introduce ω* effects. We find the diamagnetic drifts can have a stabilizing effect on the nonlinear DTM through a combination of large scale differential rotation and mechanisms local to the reconnection layer. MRC-3D is an extended MHD code based on the libMRC computational framework. It supports nonuniform grids in curvilinear coordinates with parallel implicit and explicit time integration.

Enabling co-simulation of tokamak plant models and plasma control systems

DOE PAGES

Walker, M. L.

2017-12-22

A system for connecting the Plasma Control System and a model of the tokamak Plant in closed loop co-simulation for plasma control development has been in routine use at DIII-D for more than 20 years and at other fusion labs that use variants of the DIII-D PCS for approximately the last decade. Here, co-simulation refers to the simultaneous execution of two independent codes with the exchange of data - Plant actuator commands and tokamak diagnostic data - between them during execution. Interest in this type of PCS-Plant simulation technology has also been growing recently at other fusion facilities. In fact,more » use of such closed loop control simulations is assumed to play an even larger role in the development of both the ITER Plasma Control System (PCS) and the experimental operation of the ITER device, where they will be used to support verification/validation of the PCS and also for ITER pulse schedule development and validation. We describe the key use cases that motivate the co-simulation capability and the features that must be provided by the Plasma Control System to support it. These features could be provided by the PCS itself or by a model of the PCS. If the PCS itself is chosen to provide them, there are requirements imposed on its architecture. If a PCS model is chosen, there are requirements imposed on the initial implementation of this simulation as well as long-term consequences for its continued development and maintenance. We describe these issues for each use case and discuss the relative merits of the two choices. Several examples are given illustrating uses of the co-simulation method to address problems of plasma control during the operation of DIII-D and of other devices that use the DIII-D PCS.« less

Enabling co-simulation of tokamak plant models and plasma control systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walker, M. L.

A system for connecting the Plasma Control System and a model of the tokamak Plant in closed loop co-simulation for plasma control development has been in routine use at DIII-D for more than 20 years and at other fusion labs that use variants of the DIII-D PCS for approximately the last decade. Here, co-simulation refers to the simultaneous execution of two independent codes with the exchange of data - Plant actuator commands and tokamak diagnostic data - between them during execution. Interest in this type of PCS-Plant simulation technology has also been growing recently at other fusion facilities. In fact,more » use of such closed loop control simulations is assumed to play an even larger role in the development of both the ITER Plasma Control System (PCS) and the experimental operation of the ITER device, where they will be used to support verification/validation of the PCS and also for ITER pulse schedule development and validation. We describe the key use cases that motivate the co-simulation capability and the features that must be provided by the Plasma Control System to support it. These features could be provided by the PCS itself or by a model of the PCS. If the PCS itself is chosen to provide them, there are requirements imposed on its architecture. If a PCS model is chosen, there are requirements imposed on the initial implementation of this simulation as well as long-term consequences for its continued development and maintenance. We describe these issues for each use case and discuss the relative merits of the two choices. Several examples are given illustrating uses of the co-simulation method to address problems of plasma control during the operation of DIII-D and of other devices that use the DIII-D PCS.« less

Theory and Simulations of Incomplete Reconnection During Sawteeth Due to Diamagnetic Effects

NASA Astrophysics Data System (ADS)

Beidler, Matthew Thomas

Tokamaks use magnetic fields to confine plasmas to achieve fusion; they are the leading approach proposed for the widespread production of fusion energy. The sawtooth crash in tokamaks limits the core temperature, adversely impacts confinement, and seeds disruptions. Adequate knowledge of the physics governing the sawtooth crash and a predictive capability of its ramifications has been elusive, including an understanding of incomplete reconnection, i.e., why sawteeth often cease prematurely before processing all available magnetic flux. In this dissertation, we introduce a model for incomplete reconnection in sawtooth crashes resulting from increasing diamagnetic effects in the nonlinear phase of magnetic reconnection. Physically, the reconnection inflow self-consistently convects the high pressure core of a tokamak toward the q=1 rational surface, thereby increasing the pressure gradient at the reconnection site. If the pressure gradient at the rational surface becomes large enough due to the self-consistent evolution, incomplete reconnection will occur due to diamagnetic effects becoming large enough to suppress reconnection. Predictions of this model are borne out in large-scale proof-of-principle two-fluid simulations of reconnection in a 2D slab geometry and are also consistent with data from the Mega Ampere Spherical Tokamak (MAST). Additionally, we present simulations from the 3D extended-MHD code M3D-C1 used to study the sawtooth crash in a 3D toroidal geometry for resistive-MHD and two-fluid models. This is the first study in a 3D tokamak geometry to show that the inclusion of two-fluid physics in the model equations is essential for recovering timescales more closely in line with experimental results compared to resistive-MHD and contrast the dynamics in the two models. We use a novel approach to sample the data in the plane of reconnection perpendicular to the (m,n)=(1,1) mode to carefully assess the reconnection physics. Using local measures of reconnection, we find that it is much faster in the two-fluid simulations, consistent with expectations based on global measures. By sampling data in the reconnection plane, we present the first observation of the quadrupole out-of-plane magnetic field appearing during sawtooth reconnection with the Hall term. We also explore how reconnection as viewed in the reconnection plane varies toroidally, which affects the symmetry of the reconnection geometry and the local diamagnetic effects. We expect our results to be useful for transport modeling in tokamaks, predicting energetic alpha-particle confinement, and assessing how sawteeth trigger disruptions. Since the model only depends on local diamagnetic and reconnection physics, it is machine independent, and should apply both to existing tokamaks and future ones such as ITER.

Tempest Simulations of Collisionless Damping of the Geodesic-Acoustic Mode in Edge-Plasma Pedestals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, X. Q.; Xiong, Z.; Nevins, W. M.

The fully nonlinear (full-f) four-dimensional TEMPEST gyrokinetic continuum code correctly produces the frequency and collisionless damping of geodesic-acoustic modes (GAMs) and zonal flow, with fully nonlinear Boltzmann electrons for the inverse aspect ratio {epsilon} scan and the tokamak safety factor q scan in homogeneous plasmas. TEMPEST simulations show that the GAMs exist in the edge pedestal for steep density and temperature gradients in the form of outgoing waves. The enhanced GAM damping may explain experimental beam emission spectroscopy measurements on the edge q scaling of the GAM amplitude.

Tempest Simulations of Collisionless Damping of the Geodesic-Acoustic Mode in Edge-Plasma Pedestals

NASA Astrophysics Data System (ADS)

Xu, X. Q.; Xiong, Z.; Gao, Z.; Nevins, W. M.; McKee, G. R.

2008-05-01

The fully nonlinear (full-f) four-dimensional TEMPEST gyrokinetic continuum code correctly produces the frequency and collisionless damping of geodesic-acoustic modes (GAMs) and zonal flow, with fully nonlinear Boltzmann electrons for the inverse aspect ratio γ scan and the tokamak safety factor q scan in homogeneous plasmas. TEMPEST simulations show that the GAMs exist in the edge pedestal for steep density and temperature gradients in the form of outgoing waves. The enhanced GAM damping may explain experimental beam emission spectroscopy measurements on the edge q scaling of the GAM amplitude.

TEMPEST simulations of collisionless damping of the geodesic-acoustic mode in edge-plasma pedestals.

PubMed

Xu, X Q; Xiong, Z; Gao, Z; Nevins, W M; McKee, G R

2008-05-30

The fully nonlinear (full-f) four-dimensional TEMPEST gyrokinetic continuum code correctly produces the frequency and collisionless damping of geodesic-acoustic modes (GAMs) and zonal flow, with fully nonlinear Boltzmann electrons for the inverse aspect ratio scan and the tokamak safety factor q scan in homogeneous plasmas. TEMPEST simulations show that the GAMs exist in the edge pedestal for steep density and temperature gradients in the form of outgoing waves. The enhanced GAM damping may explain experimental beam emission spectroscopy measurements on the edge q scaling of the GAM amplitude.

Flux tube gyrokinetic simulations of the edge pedestal

NASA Astrophysics Data System (ADS)

Parker, Scott; Wan, Weigang; Chen, Yang

2011-10-01

The linear instabilities of DIII-D H-mode pedestal are studied with gyrokinetic micro-turbulence simulations. The simulation code GEM is an electromagnetic δf code with global tokamak geometry in the form of Miller equilibrium. Local flux tube simulations are carried out for multiple positions of two DIII-D profiles: shot #98889 and shot #131997. Near the top of the pedestal, the instability is clearly ITG. The dominant instability of the pedestal appears at the steep gradient region, and it is identified as a low frequency mode mostly driven by electron temperature gradient. The mode propagates along the electron diamagnetic direction for low n and may propagate along the ion direction for high n. At some positions near the steep gradient region, an ion instability is found which shows some characteristics of kinetic ballooning mode (KBM). These results will be compared to the results of E. Wang et al. and D. Fulton et al. in the same session. We thank R. Groebner and P. Snyder for providing experimental profiles and helpful discussions.

CXSFIT Code Application to Process Charge-Exchange Recombination Spectroscopy Data at the T-10 Tokamak

NASA Astrophysics Data System (ADS)

Serov, S. V.; Tugarinov, S. N.; Klyuchnikov, L. A.; Krupin, V. A.; von Hellermann, M.

2017-12-01

The applicability of the CXSFIT code to process experimental data from Charge-eXchange Recombination Spectroscopy (CXRS) diagnostics at the T-10 tokamak is studied with a view to its further use for processing experimental data at the ITER facility. The design and operating principle of the CXRS diagnostics are described. The main methods for processing the CXRS spectra of the 5291-Å line of C5+ ions at the T-10 tokamak (with and without subtraction of parasitic emission from the edge plasma) are analyzed. The method of averaging the CXRS spectra over several shots, which is used at the T-10 tokamak to increase the signal-to-noise ratio, is described. The approximation of the spectrum by a set of Gaussian components is used to identify the active CXRS line in the measured spectrum. Using the CXSFIT code, the ion temperature in ohmic discharges and discharges with auxiliary electron cyclotron resonance heating (ECRH) at the T-10 tokamak is calculated from the CXRS spectra of the 5291-Å line. The time behavior of the ion temperature profile in different ohmic heating modes is studied. The temperature profile dependence on the ECRH power is measured, and the dynamics of ECR removal of carbon nuclei from the T-10 plasma is described. Experimental data from the CXRS diagnostics at T-10 substantially contribute to the implementation of physical programs of studies on heat and particle transport in tokamak plasmas and investigation of geodesic acoustic mode properties.

Disruption forces on the tokamak wall with and without poloidal currents

NASA Astrophysics Data System (ADS)

Pustovitov, V. D.

2017-05-01

The contributions into the disruption radial force on the tokamak vacuum vessel wall are calculated and analyzed. One is due to the induced toroidal current in the wall, and another is due to the poloidal current. The latter is not accounted for in the models that represent the wall as a set of isolated toroidal filaments. It is shown that such modeling must lead to significant errors in the evaluation of the force during either thermal or current quench. The analytical derivations are performed here for an arbitrary tokamak configuration with final estimates for a circular large-aspect-ratio plasma and a coaxial wall reacting on perturbations as a perfect conductor. The results are compared with those recently obtained numerically by the codes DINA, MAXFEA and CarMa0NL. The discrepancies between the DINA simulations (Khayrutdinov et al 2016 Plasma Phys. Control. Fusion 58 115012) and earlier analytical predictions are explained. The recent conclusion (Villone et al 2015 Fusion Eng. Des. 93 57) on the role of the disruption-induced poloidal current in the wall is confirmed and extended to a wider area.

Final Technical Report for SBIR entitled Four-Dimensional Finite-Orbit-Width Fokker-Planck Code with Sources, for Neoclassical/Anomalous Transport Simulation of Ion and Electron Distributions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harvey, R. W.; Petrov, Yu. V.

2013-12-03

Within the US Department of Energy/Office of Fusion Energy magnetic fusion research program, there is an important whole-plasma-modeling need for a radio-frequency/neutral-beam-injection (RF/NBI) transport-oriented finite-difference Fokker-Planck (FP) code with combined capabilities for 4D (2R2V) geometry near the fusion plasma periphery, and computationally less demanding 3D (1R2V) bounce-averaged capabilities for plasma in the core of fusion devices. Demonstration of proof-of-principle achievement of this goal has been carried out in research carried out under Phase I of the SBIR award. Two DOE-sponsored codes, the CQL3D bounce-average Fokker-Planck code in which CompX has specialized, and the COGENT 4D, plasma edge-oriented Fokker-Planck code whichmore » has been constructed by Lawrence Livermore National Laboratory and Lawrence Berkeley Laboratory scientists, where coupled. Coupling was achieved by using CQL3D calculated velocity distributions including an energetic tail resulting from NBI, as boundary conditions for the COGENT code over the two-dimensional velocity space on a spatial interface (flux) surface at a given radius near the plasma periphery. The finite-orbit-width fast ions from the CQL3D distributions penetrated into the peripheral plasma modeled by the COGENT code. This combined code demonstrates the feasibility of the proposed 3D/4D code. By combining these codes, the greatest computational efficiency is achieved subject to present modeling needs in toroidally symmetric magnetic fusion devices. The more efficient 3D code can be used in its regions of applicability, coupled to the more computationally demanding 4D code in higher collisionality edge plasma regions where that extended capability is necessary for accurate representation of the plasma. More efficient code leads to greater use and utility of the model. An ancillary aim of the project is to make the combined 3D/4D code user friendly. Achievement of full-coupling of these two Fokker-Planck codes will advance computational modeling of plasma devices important to the USDOE magnetic fusion energy program, in particular the DIII-D tokamak at General Atomics, San Diego, the NSTX spherical tokamak at Princeton, New Jersey, and the MST reversed-field-pinch Madison, Wisconsin. The validation studies of the code against the experiments will improve understanding of physics important for magnetic fusion, and will increase our design capabilities for achieving the goals of the International Tokamak Experimental Reactor (ITER) project in which the US is a participant and which seeks to demonstrate at least a factor of five in fusion power production divided by input power.« less

Numerical optimization of perturbative coils for tokamaks

NASA Astrophysics Data System (ADS)

Lazerson, Samuel; Park, Jong-Kyu; Logan, Nikolas; Boozer, Allen; NSTX-U Research Team

2014-10-01

Numerical optimization of coils which apply three dimensional (3D) perturbative fields to tokamaks is presented. The application of perturbative 3D magnetic fields in tokamaks is now commonplace for control of error fields, resistive wall modes, resonant field drive, and neoclassical toroidal viscosity (NTV) torques. The design of such systems has focused on control of toroidal mode number, with coil shapes based on simple window-pane designs. In this work, a numerical optimization suite based on the STELLOPT 3D equilibrium optimization code is presented. The new code, IPECOPT, replaces the VMEC equilibrium code with the IPEC perturbed equilibrium code, and targets NTV torque by coupling to the PENT code. Fixed boundary optimizations of the 3D fields for the NSTX-U experiment are underway. Initial results suggest NTV torques can be driven by normal field spectrums which are not pitch-resonant with the magnetic field lines. Work has focused on driving core torque with n = 1 and edge torques with n = 3 fields. Optimizations of the coil currents for the planned NSTX-U NCC coils highlight the code's free boundary capability. This manuscript has been authored by Princeton University under Contract Number DE-AC02-09CH11466 with the U.S. Department of Energy.

Saturated Widths of Magnetic Islands in Tokamak Discharges

NASA Astrophysics Data System (ADS)

Halpern, F.; Pankin, A. Y.

2005-10-01

The new ISLAND module described in reference [1] implements a quasi-linear model to compute the widths of multiple magnetic islands driven by saturated tearing modes in toroidal plasmas of arbitrary aspect ratio and cross sectional shape. The distortion of the island shape caused by the radial variation in the perturbation is computed in the new module. In transport simulations, the enhanced transport caused by the magnetic islands has the effect of flattening the pressure and current density profiles. This self consistent treatment of the magnetic islands alters the development of the plasma profiles. In addition, it is found that islands closer to the magnetic axis influence the evolution of islands further out in the plasma. In order to investigate such phenomena, the ISLAND module is used within the BALDUR predictive modeling code to compute the widths of multiple magnetic islands in tokamak discharges. The interaction between the islands and sawtooth crashes is examined in simulations of DIII-D and JET discharges. The module is used to compute saturated neoclassical tearing mode island widths for multiple modes in ITER. Preliminary results for island widths in ITER are consistent with those presented [2] by Hegna. [1] F.D. Halpern, G. Bateman, A.H. Kritz and A.Y. Pankin, ``The ISLAND Module for Computing Magnetic Island Widths in Tokamaks,'' submitted to J. Plasma Physics (2005). [2] C.C. Hegna, 2002 Fusion Snowmass Meeting.

Gyrokinetic-Vlasov simulations of the ion temperature gradient turbulence in tokamak and helical systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Watanabe, T.-H.; Sugama, H.; Graduate University for Advanced Studies

2006-11-30

Recent progress of the gyrokinetic-Vlasov simulations on the ion temperature gradient (ITG) turbulence in tokamak and helical systems is reported, where the entropy balance is checked as a reference for the numerical accuracy. The tokamak ITG turbulence simulation carried out on the Earth Simulator clearly captures a nonlinear generation process of zonal flows. The tera-flops and tera-bytes scale simulation is also applied to a helical system with the same poloidal and toroidal periodicities of L = 2 and M = 10 as in the Large Helical Device.

Analysis of JT-60SA operational scenarios

NASA Astrophysics Data System (ADS)

Garzotti, L.; Barbato, E.; Garcia, J.; Hayashi, N.; Voitsekhovitch, I.; Giruzzi, G.; Maget, P.; Romanelli, M.; Saarelma, S.; Stankiewitz, R.; Yoshida, M.; Zagórski, R.

2018-02-01

Reference scenarios for the JT-60SA tokamak have been simulated with one-dimensional transport codes to assess the stationary state of the flat-top phase and provide a profile database for further physics studies (e.g. MHD stability, gyrokinetic analysis) and diagnostics design. The types of scenario considered vary from pulsed standard H-mode to advanced non-inductive steady-state plasmas. In this paper we present the results obtained with the ASTRA, CRONOS, JINTRAC and TOPICS codes equipped with the Bohm/gyro-Bohm, CDBM and GLF23 transport models. The scenarios analysed here are: a standard ELMy H-mode, a hybrid scenario and a non-inductive steady state plasma, with operational parameters from the JT-60SA research plan. Several simulations of the scenarios under consideration have been performed with the above mentioned codes and transport models. The results from the different codes are in broad agreement and the main plasma parameters generally agree well with the zero dimensional estimates reported previously. The sensitivity of the results to different transport models and, in some cases, to the ELM/pedestal model has been investigated.

Hybrid simulation of fishbone instabilities in the EAST tokamak

DOE PAGES

Shen, Wei; Wang, Feng; Fu, G. Y.; ...

2017-08-11

Hybrid simulations with the global kinetic-magnetohydrodynamic (MHD) code M3D-K have been carried out to investigate the linear stability and nonlinear dynamics of beam-driven fishbone in the experimental advanced superconducting tokamak (EAST) experiment. Linear simulations show that a low frequency fishbone instability is excited at experimental value of beam ion pressure. The mode is mainly driven by low energy beam ions via precessional resonance. Our results are consistent with the experimental measurement with respect to mode frequency and mode structure. When the beam ion pressure is increased to exceed a critical value, the low frequency mode transits to a beta-induced Alfvenmore » eigenmode (BAE) with much higher frequency. This BAE is driven by higher energy beam ions. Nonlinear simulations show that the frequency of the low frequency fishbone chirps up and down with corresponding hole-clump structures in phase space, consistent with the Berk-Breizman theory. In addition to the low frequency mode, the high frequency BAE is excited during the nonlinear evolution. Furthermore, for the transient case of beam pressure fraction where the low and high frequency modes are simultaneously excited in the linear phase, only one dominant mode appears in the nonlinear phase with frequency jumps up and down during nonlinear evolution.« less

Comparisons between tokamak fueling of gas puffing and supersonic molecular beam injection in 2D simulations

DOE PAGES

Zhou, Y. L.; Wang, Z. H.; Xu, X. Q.; ...

2015-01-09

Plasma fueling with high efficiency and deep injection is very important to enable fusion power performance requirements. It is a powerful and efficient way to study neutral transport dynamics and find methods of improving the fueling performance by doing large scale simulations. Furthermore, two basic fueling methods, gas puffing (GP) and supersonic molecular beam injection (SMBI), are simulated and compared in realistic divertor geometry of the HL-2A tokamak with a newly developed module, named trans-neut, within the framework of BOUT++ boundary plasma turbulence code [Z. H. Wang et al., Nucl. Fusion 54, 043019 (2014)]. The physical model includes plasma density,more » heat and momentum transport equations along with neutral density, and momentum transport equations. In transport dynamics and profile evolutions of both plasma and neutrals are simulated and compared between GP and SMBI in both poloidal and radial directions, which are quite different from one and the other. It finds that the neutrals can penetrate about four centimeters inside the last closed (magnetic) flux surface during SMBI, while they are all deposited outside of the LCF during GP. Moreover, it is the radial convection and larger inflowing flux which lead to the deeper penetration depth of SMBI and higher fueling efficiency compared to GP.« less

Comparisons between tokamak fueling of gas puffing and supersonic molecular beam injection in 2D simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhou, Y. L.; Wang, Z. H.; Xu, X. Q.

Plasma fueling with high efficiency and deep injection is very important to enable fusion power performance requirements. It is a powerful and efficient way to study neutral transport dynamics and find methods of improving the fueling performance by doing large scale simulations. Furthermore, two basic fueling methods, gas puffing (GP) and supersonic molecular beam injection (SMBI), are simulated and compared in realistic divertor geometry of the HL-2A tokamak with a newly developed module, named trans-neut, within the framework of BOUT++ boundary plasma turbulence code [Z. H. Wang et al., Nucl. Fusion 54, 043019 (2014)]. The physical model includes plasma density,more » heat and momentum transport equations along with neutral density, and momentum transport equations. In transport dynamics and profile evolutions of both plasma and neutrals are simulated and compared between GP and SMBI in both poloidal and radial directions, which are quite different from one and the other. It finds that the neutrals can penetrate about four centimeters inside the last closed (magnetic) flux surface during SMBI, while they are all deposited outside of the LCF during GP. Moreover, it is the radial convection and larger inflowing flux which lead to the deeper penetration depth of SMBI and higher fueling efficiency compared to GP.« less

Comparisons between tokamak fueling of gas puffing and supersonic molecular beam injection in 2D simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhou, Y. L.; Southwestern Institute of Physics, Chengdu 610041; Wang, Z. H., E-mail: zhwang@swip.ac.cn

Plasma fueling with high efficiency and deep injection is very important to enable fusion power performance requirements. It is a powerful and efficient way to study neutral transport dynamics and find methods of improving the fueling performance by doing large scale simulations. Two basic fueling methods, gas puffing (GP) and supersonic molecular beam injection (SMBI), are simulated and compared in realistic divertor geometry of the HL-2A tokamak with a newly developed module, named trans-neut, within the framework of BOUT++ boundary plasma turbulence code [Z. H. Wang et al., Nucl. Fusion 54, 043019 (2014)]. The physical model includes plasma density, heatmore » and momentum transport equations along with neutral density, and momentum transport equations. Transport dynamics and profile evolutions of both plasma and neutrals are simulated and compared between GP and SMBI in both poloidal and radial directions, which are quite different from one and the other. It finds that the neutrals can penetrate about four centimeters inside the last closed (magnetic) flux surface during SMBI, while they are all deposited outside of the LCF during GP. It is the radial convection and larger inflowing flux which lead to the deeper penetration depth of SMBI and higher fueling efficiency compared to GP.« less

Status of BOUT fluid turbulence code: improvements and verification

NASA Astrophysics Data System (ADS)

Umansky, M. V.; Lodestro, L. L.; Xu, X. Q.

2006-10-01

BOUT is an electromagnetic fluid turbulence code for tokamak edge plasma [1]. BOUT performs time integration of reduced Braginskii plasma fluid equations, using spatial discretization in realistic geometry and employing a standard ODE integration package PVODE. BOUT has been applied to several tokamak experiments and in some cases calculated spectra of turbulent fluctuations compared favorably to experimental data. On the other hand, the desire to understand better the code results and to gain more confidence in it motivated investing effort in rigorous verification of BOUT. Parallel to the testing the code underwent substantial modification, mainly to improve its readability and tractability of physical terms, with some algorithmic improvements as well. In the verification process, a series of linear and nonlinear test problems was applied to BOUT, targeting different subgroups of physical terms. The tests include reproducing basic electrostatic and electromagnetic plasma modes in simplified geometry, axisymmetric benchmarks against the 2D edge code UEDGE in real divertor geometry, and neutral fluid benchmarks against the hydrodynamic code LCPFCT. After completion of the testing, the new version of the code is being applied to actual tokamak edge turbulence problems, and the results will be presented. [1] X. Q. Xu et al., Contr. Plas. Phys., 36,158 (1998). *Work performed for USDOE by Univ. Calif. LLNL under contract W-7405-ENG-48.

Studies of numerical algorithms for gyrokinetics and the effects of shaping on plasma turbulence

NASA Astrophysics Data System (ADS)

Belli, Emily Ann

Advanced numerical algorithms for gyrokinetic simulations are explored for more effective studies of plasma turbulent transport. The gyrokinetic equations describe the dynamics of particles in 5-dimensional phase space, averaging over the fast gyromotion, and provide a foundation for studying plasma microturbulence in fusion devices and in astrophysical plasmas. Several algorithms for Eulerian/continuum gyrokinetic solvers are compared. An iterative implicit scheme based on numerical approximations of the plasma response is developed. This method reduces the long time needed to set-up implicit arrays, yet still has larger time step advantages similar to a fully implicit method. Various model preconditioners and iteration schemes, including Krylov-based solvers, are explored. An Alternating Direction Implicit algorithm is also studied and is surprisingly found to yield a severe stability restriction on the time step. Overall, an iterative Krylov algorithm might be the best approach for extensions of core tokamak gyrokinetic simulations to edge kinetic formulations and may be particularly useful for studies of large-scale ExB shear effects. The effects of flux surface shape on the gyrokinetic stability and transport of tokamak plasmas are studied using the nonlinear GS2 gyrokinetic code with analytic equilibria based on interpolations of representative JET-like shapes. High shaping is found to be a stabilizing influence on both the linear ITG instability and nonlinear ITG turbulence. A scaling of the heat flux with elongation of chi ˜ kappa-1.5 or kappa-2 (depending on the triangularity) is observed, which is consistent with previous gyrofluid simulations. Thus, the GS2 turbulence simulations are explaining a significant fraction, but not all, of the empirical elongation scaling. The remainder of the scaling may come from (1) the edge boundary conditions for core turbulence, and (2) the larger Dimits nonlinear critical temperature gradient shift due to the enhancement of zonal flows with shaping, which is observed with the GS2 simulations. Finally, a local linear trial function-based gyrokinetic code is developed to aid in fast scoping studies of gyrokinetic linear stability. This code is successfully benchmarked with the full GS2 code in the collisionless, electrostatic limit, as well as in the more general electromagnetic description with higher-order Hermite basis functions.

Simulation of profile evolution from ramp-up to ramp-down and optimization of tokamak plasma termination with the RAPTOR code

NASA Astrophysics Data System (ADS)

Teplukhina, A. A.; Sauter, O.; Felici, F.; Merle, A.; Kim, D.; the TCV Team; the ASDEX Upgrade Team; the EUROfusion MST1 Team

2017-12-01

The present work demonstrates the capabilities of the transport code RAPTOR as a fast and reliable simulator of plasma profiles for the entire plasma discharge, i.e. from ramp-up to ramp-down. This code focuses, at this stage, on the simulation of electron temperature and poloidal flux profiles using prescribed equilibrium and some kinetic profiles. In this work we extend the RAPTOR transport model to include a time-varying plasma equilibrium geometry and verify the changes via comparison with ATSRA code simulations. In addition a new ad hoc transport model based on constant gradients and suitable for simulations of L-H and H-L mode transitions has been incorporated into the RAPTOR code and validated with rapid simulations of the time evolution of the safety factor and the electron temperature over the entire AUG and TCV discharges. An optimization procedure for the plasma termination phase has also been developed during this work. We define the goal of the optimization as ramping down the plasma current as fast as possible while avoiding any disruptions caused by reaching physical or technical limits. Our numerical study of this problem shows that a fast decrease of plasma elongation during current ramp-down can help in reducing plasma internal inductance. An early transition from H- to L-mode allows us to reduce the drop in poloidal beta, which is also important for plasma MHD stability and control. This work shows how these complex nonlinear interactions can be optimized automatically using relevant cost functions and constraints. Preliminary experimental results for TCV are demonstrated.

Gyrokinetic continuum simulations of turbulence in the Texas Helimak

NASA Astrophysics Data System (ADS)

Bernard, T. N.; Shi, E. L.; Hammett, G. W.; Hakim, A.; Taylor, E. I.

2017-10-01

We have used the Gkeyll code to perform 3x-2v full-f gyrokinetic continuum simulations of electrostatic plasma turbulence in the Texas Helimak. The Helimak is an open field-line experiment with magnetic curvature and shear. It is useful for validating numerical codes due to its extensive diagnostics and simple, helical geometry, which is similar to the scrape-off layer region of tokamaks. Interchange and drift-wave modes are the main turbulence mechanisms in the device, and potential biasing is applied to study the effect of velocity shear on turbulence reduction. With Gkeyll, we varied field-line pitch angle and simulated biased and unbiased cases to study different turbulent regimes and turbulence reduction. These are the first kinetic simulations of the Helimak and resulting plasma profiles agree fairly well with experimental data. This research demonstrates Gkeyll's progress towards 5D simulations of the SOL region of fusion devices. Supported by the U.S. DOE SCGSR program under contract DE-SC0014664, the Max-Planck/Princeton Center for Plasma Physics, the SciDAC Center for the Study of Plasma Microturbulence, and DOE contract DE-AC02-09CH11466.

Modeling of Resistive Wall Modes in Tokamak and Reversed Field Pinch Configurations of KTX

NASA Astrophysics Data System (ADS)

Han, Rui; Zhu, Ping; Bai, Wei; Lan, Tao; Liu, Wandong

2016-10-01

Resistive wall mode is believed to be one of the leading causes for macroscopic degradation of plasma confinement in tokamaks and reversed field pinches (RFP). In this study, we evaluate the linear RWM instability of Keda Torus eXperiment (KTX) in both tokamak and RFP configurations. For the tokamak configuration, the extended MHD code NIMROD is employed for calculating the dependence of the RWM growth rate on the position and conductivity of the vacuum wall for a model tokamak equilibrium of KTX in the large aspect-ratio approximation. For the RFP configuration, the standard formulation of dispersion relation for RWM based on the MHD energy principle has been evaluated for a cylindrical α- Θ model of KTX plasma equilibrium, in an effort to investigate the effects of thin wall on the RWM in KTX. Full MHD calculations of RWM in the RFP configuration of KTX using the NIMROD code are also being developed. Supported by National Magnetic Confinement Fusion Science Program of China Grant Nos. 2014GB124002, 2015GB101004, 2011GB106000, and 2011GB106003.

Numerical studies of edge localized instabilities in tokamaks

NASA Astrophysics Data System (ADS)

Wilson, H. R.; Snyder, P. B.; Huysmans, G. T. A.; Miller, R. L.

2002-04-01

A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code.

The impact of collisionality, FLR, and parallel closure effects on instabilities in the tokamak pedestal: Numerical studies with the NIMROD code

DOE PAGES

King, J. R.; Pankin, A. Y.; Kruger, S. E.; ...

2016-06-24

The extended-MHD NIMROD code [C. R. Sovinec and J. R. King, J. Comput. Phys. 229, 5803 (2010)] is verified against the ideal-MHD ELITE code [H. R. Wilson et al., Phys. Plasmas 9, 1277 (2002)] on a diverted tokamak discharge. When the NIMROD model complexity is increased incrementally, resistive and first-order finite-Larmour radius effects are destabilizing and stabilizing, respectively. Lastly, the full result is compared to local analytic calculations which are found to overpredict both the resistive destabilization and drift stabilization in comparison to the NIMROD computations.

The impact of collisionality, FLR, and parallel closure effects on instabilities in the tokamak pedestal: Numerical studies with the NIMROD code

DOE Office of Scientific and Technical Information (OSTI.GOV)

King, J. R.; Pankin, A. Y.; Kruger, S. E.

The extended-MHD NIMROD code [C. R. Sovinec and J. R. King, J. Comput. Phys. 229, 5803 (2010)] is verified against the ideal-MHD ELITE code [H. R. Wilson et al., Phys. Plasmas 9, 1277 (2002)] on a diverted tokamak discharge. When the NIMROD model complexity is increased incrementally, resistive and first-order finite-Larmour radius effects are destabilizing and stabilizing, respectively. The full result is compared to local analytic calculations which are found to overpredict both the resistive destabilization and drift stabilization in comparison to the NIMROD computations.

The impact of collisionality, FLR, and parallel closure effects on instabilities in the tokamak pedestal: Numerical studies with the NIMROD code

DOE Office of Scientific and Technical Information (OSTI.GOV)

King, J. R.; Pankin, A. Y.; Kruger, S. E.

The extended-MHD NIMROD code [C. R. Sovinec and J. R. King, J. Comput. Phys. 229, 5803 (2010)] is verified against the ideal-MHD ELITE code [H. R. Wilson et al., Phys. Plasmas 9, 1277 (2002)] on a diverted tokamak discharge. When the NIMROD model complexity is increased incrementally, resistive and first-order finite-Larmour radius effects are destabilizing and stabilizing, respectively. Lastly, the full result is compared to local analytic calculations which are found to overpredict both the resistive destabilization and drift stabilization in comparison to the NIMROD computations.

A fast low-to-high confinement mode bifurcation dynamics in the boundary-plasma gyrokinetic code XGC1

NASA Astrophysics Data System (ADS)

Ku, S.; Chang, C. S.; Hager, R.; Churchill, R. M.; Tynan, G. R.; Cziegler, I.; Greenwald, M.; Hughes, J.; Parker, S. E.; Adams, M. F.; D'Azevedo, E.; Worley, P.

2018-05-01

A fast edge turbulence suppression event has been simulated in the electrostatic version of the gyrokinetic particle-in-cell code XGC1 in a realistic diverted tokamak edge geometry under neutral particle recycling. The results show that the sequence of turbulent Reynolds stress followed by neoclassical ion orbit-loss driven together conspire to form the sustaining radial electric field shear and to quench turbulent transport just inside the last closed magnetic flux surface. The main suppression action is located in a thin radial layer around ψN≃0.96 -0.98 , where ψN is the normalized poloidal flux, with the time scale ˜0.1 ms.

TEMPEST Simulations of Collisionless Damping of Geodesic-Acoustic Mode in Edge Plasma Pedestal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, X Q; Xiong, Z; Nevins, W M

The fully nonlinear (full-f) 4D TEMPEST gyrokinetic continuum code produces frequency, collisionless damping of GAM and zonal flow with fully nonlinear Boltzmann electrons for the inverse aspect ratio {epsilon}-scan and the tokamak safety factor q-scan in homogeneous plasmas. The TEMPEST simulation shows that GAM exists in edge plasma pedestal for steep density and temperature gradients, and an initial GAM relaxes to the standard neoclassical residual, rather than Rosenbluth-Hinton residual due to the presence of ion-ion collisions. The enhanced GAM damping explains experimental BES measurements on the edge q scaling of the GAM amplitude.

TEMPEST Simulations of Collisionless Damping of Geodesic-Acoustic Mode in Edge Plasma Pedestal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, X; Xiong, Z; Nevins, W

The fully nonlinear 4D TEMPEST gyrokinetic continuum code produces frequency, collisionless damping of geodesic-acoustic mode (GAM) and zonal flow with fully nonlinear Boltzmann electrons for the inverse aspect ratio {epsilon}-scan and the tokamak safety factor q-scan in homogeneous plasmas. The TEMPEST simulation shows that GAM exists in edge plasma pedestal for steep density and temperature gradients, and an initial GAM relaxes to the standard neoclassical residual, rather than Rosenbluth-Hinton residual due to the presence of ion-ion collisions. The enhanced GAM damping explains experimental BES measurements on the edge q scaling of the GAM amplitude.

Calculations of Helium Bubble Evolution in the PISCES Experiments with Cluster Dynamics

NASA Astrophysics Data System (ADS)

Blondel, Sophie; Younkin, Timothy; Wirth, Brian; Lasa, Ane; Green, David; Canik, John; Drobny, Jon; Curreli, Davide

2017-10-01

Plasma surface interactions in fusion tokamak reactors involve an inherently multiscale, highly non-equilibrium set of phenomena, for which current models are inadequate to predict the divertor response to and feedback on the plasma. In this presentation, we describe the latest code developments of Xolotl, a spatially-dependent reaction diffusion cluster dynamics code to simulate the divertor surface response to fusion-relevant plasma exposure. Xolotl is part of a code-coupling effort to model both plasma and material simultaneously; the first benchmark for this effort is the series of PISCES linear device experiments. We will discuss the processes leading to surface morphology changes, which further affect erosion, as well as how Xolotl has been updated in order to communicate with other codes. Furthermore, we will show results of the sub-surface evolution of helium bubbles in tungsten as well as the material surface displacement under these conditions.

Versatile fusion source integrator AFSI for fast ion and neutron studies in fusion devices

NASA Astrophysics Data System (ADS)

Sirén, Paula; Varje, Jari; Äkäslompolo, Simppa; Asunta, Otto; Giroud, Carine; Kurki-Suonio, Taina; Weisen, Henri; JET Contributors, The

2018-01-01

ASCOT Fusion Source Integrator AFSI, an efficient tool for calculating fusion reaction rates and characterizing the fusion products, based on arbitrary reactant distributions, has been developed and is reported in this paper. Calculation of reactor-relevant D-D, D-T and D-3He fusion reactions has been implemented based on the Bosch-Hale fusion cross sections. The reactions can be calculated between arbitrary particle populations, including Maxwellian thermal particles and minority energetic particles. Reaction rate profiles, energy spectra and full 4D phase space distributions can be calculated for the non-isotropic reaction products. The code is especially suitable for integrated modelling in self-consistent plasma physics simulations as well as in the Serpent neutronics calculation chain. Validation of the model has been performed for neutron measurements at the JET tokamak and the code has been applied to predictive simulations in ITER.

Feasibility of a motional Stark effect system on the TCV tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siegrist, M.R.; Hawkes, N.; Weisen, H.

This paper presents a feasibility study for a motional Stark effect (MSE) [F. M. Levinton et al., Phys. Rev. Lett. 63, 2060 (1989)] diagnostic on the TCV tokamak. A numerical simulation code has been used to identify the optimal port arrangement and geometrical layout. It predicts the expected measurement accuracy for a range of typical plasma scenarios. With the existing neutral beam injector (NBI) and a detection system based on current day technology, it should be possible to determine the safety factor with an accuracy of the order of 5%. A vertically injected beam through the plasma center would allowmore » one to measure plasmas which are centered above the midplane, a common occurrence in connection with electron cyclotron resonance heating and electron cyclotron current drive experiments. In this case a new and ideally more powerful NBI would be required.« less

Hα line shape in front of the limiter in the HT-6M tokamak

NASA Astrophysics Data System (ADS)

Wan, Baonian; Li, Jiangang; Luo, Jiarong; Xie, Jikang; Wu, Zhenwei; Zhang, Xianmei; HT-6M Group

1999-11-01

The Hα line shape in front of the limiter in the HT-6M tokamak is analysed by multi-Gaussian fitting. The energy distribution of neutral hydrogen atoms reveals that Hα radiation is contributed by Franck-Condon atoms, atoms reflected at the limiter surface and charge exchange. Multi-Gaussian fitting of the Hα spectral profile indicates contributions of 60% from reflection particles and 40% from molecule dissociation to recycling. Ion temperatures in central regions are obtained from the spectral width of charge exchange components. Dissociation of hydrogen molecules and reflection of particles at the limiter surface are dominant in edge recycling. Reduction of particle reflection at the limiter surface is important for controlling edge recycling. The measured profiles of neutral hydrogen atom density are reproduced by a particle continuity equation and a simplified one dimensional Monte Carlo simulation code.

Tokamak plasma high field side response to an n = 3 magnetic perturbation: a comparison of 3D equilibrium solutions from seven different codes

NASA Astrophysics Data System (ADS)

Reiman, A.; Ferraro, N. M.; Turnbull, A.; Park, J. K.; Cerfon, A.; Evans, T. E.; Lanctot, M. J.; Lazarus, E. A.; Liu, Y.; McFadden, G.; Monticello, D.; Suzuki, Y.

2015-06-01

In comparing equilibrium solutions for a DIII-D shot that is amenable to analysis by both stellarator and tokamak three-dimensional (3D) equilibrium codes, a significant disagreement has been seen between solutions of the VMEC stellarator equilibrium code and solutions of tokamak perturbative 3D equilibrium codes. The source of that disagreement has been investigated, and that investigation has led to new insights into the domain of validity of the different equilibrium calculations, and to a finding that the manner in which localized screening currents at low order rational surfaces are handled can affect global properties of the equilibrium solution. The perturbative treatment has been found to break down at surprisingly small perturbation amplitudes due to overlap of the calculated perturbed flux surfaces, and that treatment is not valid in the pedestal region of the DIII-D shot studied. The perturbative treatment is valid, however, further into the interior of the plasma, and flux surface overlap does not account for the disagreement investigated here. Calculated equilibrium solutions for simple model cases and comparison of the 3D equilibrium solutions with those of other codes indicate that the disagreement arises from a difference in handling of localized currents at low order rational surfaces, with such currents being absent in VMEC and present in the perturbative codes. The significant differences in the global equilibrium solutions associated with the presence or absence of very localized screening currents at rational surfaces suggests that it may be possible to extract information about localized currents from appropriate measurements of global equilibrium plasma properties. That would require improved diagnostic capability on the high field side of the tokamak plasma, a region difficult to access with diagnostics.

Nanoparticle Plasma Jet as Fast Probe for Runaway Electrons in Tokamak Disruptions

NASA Astrophysics Data System (ADS)

Bogatu, I. N.; Galkin, S. A.

2017-10-01

Successful probing of runaway electrons (REs) requires fast (1 - 2 ms) high-speed injection of enough mass able to penetrate through tokamak toroidal B-field (2 - 5 T) over 1 - 2 m distance with large assimilation fraction in core plasma. A nanoparticle plasma jet (NPPJ) from a plasma gun is a unique combination of millisecond trigger-to-delivery response and mass-velocity of 100 mg at several km/s for deep direct injection into current channel of rapidly ( 1 ms) cooling post-TQ core plasma. After C60 NPPJ test bed demonstration we started to work on ITER-compatible boron nitride (BN) NPPJ. Once injected into plasma, BN NP undergoes ablative sublimation, thermally decomposes into B and N, and releases abundant B and N high-charge ions along plasma-traversing path and into the core. We present basic characteristics of our BN NPPJ concept and first results from B and N ions on Zeff > 1 effect on REs dynamics by using a self-consistent model for RE current density. Simulation results of BNQ+ NPPJ penetration through tokamak B-field to RE beam location performed with Hybrid Electro-Magnetic code (HEM-2D) are also presented. Work supported by U.S. DOE SBIR Grant.

Numerical modeling of lower hybrid current drive in fully non-inductive plasma start-up experiments on TST-2

NASA Astrophysics Data System (ADS)

Tsujii, N.; Takase, Y.; Ejiri, A.; Shinya, T.; Togashi, H.; Yajima, S.; Yamazaki, H.; Moeller, C. P.; Roidl, B.; Sonehara, M.; Takahashi, W.; Toida, K.; Yoshida, Y.

2017-12-01

Non-inductive plasma start-up is a critical issue for spherical tokamaks since there is not enough room to provide neutron shielding for the center solenoid. Start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Because of the low magnetic field of a spherical tokamak, the plasma density needs to be kept at a very low value during the plasma current ramp-up so that the plasma core remains accessible to the LH waves. However, we have found that higher density was required to sustain larger plasma current. The achievable plasma current was limited by the maximum operational toroidal field of TST-2. The existence of an optimum density for LH current drive and its toroidal field dependence is explained through a numerical simulation based on a ray tracing code and a Fokker-Planck solver. In order to access higher density at the same magnetic field, a top-launch antenna was recently installed in addition to the existing outboard-launch antenna. Increase in the density limit was observed when the power was launched from the top antenna, consistently with the numerical predictions.

Analysis of activation and shutdown contact dose rate for EAST neutral beam port

NASA Astrophysics Data System (ADS)

Chen, Yuqing; Wang, Ji; Zhong, Guoqiang; Li, Jun; Wang, Jinfang; Xie, Yahong; Wu, Bin; Hu, Chundong

2017-12-01

For the safe operation and maintenance of neutral beam injector (NBI), specific activity and shutdown contact dose rate of the sample material SS316 are estimated around the experimental advanced superconducting tokamak (EAST) neutral beam port. Firstly, the neutron emission intensity is calculated by TRANSP code while the neutral beam is co-injected to EAST. Secondly, the neutron activation and shutdown contact dose rates for the neutral beam sample materials SS316 are derived by the Monte Carlo code MCNP and the inventory code FISPACT-2007. The simulations indicate that the primary radioactive nuclides of SS316 are 58Co and 54Mn. The peak contact dose rate is 8.52 × 10-6 Sv/h after EAST shutdown one second. That is under the International Thermonuclear Experimental Reactor (ITER) design values 1 × 10-5 Sv/h.

Progress with the COGENT Edge Kinetic Code: Collision operator options

DOE PAGES

Dorf, M. A.; Cohen, R. H.; Compton, J. C.; ...

2012-06-27

In this study, COGENT is a continuum gyrokinetic code for edge plasmas being developed by the Edge Simulation Laboratory collaboration. The code is distinguished by application of the fourth order conservative discretization, and mapped multiblock grid technology to handle the geometric complexity of the tokamak edge. It is written in v∥-μ (parallel velocity – magnetic moment) velocity coordinates, and making use of the gyrokinetic Poisson equation for the calculation of a self-consistent electric potential. In the present manuscript we report on the implementation and initial testing of a succession of increasingly detailed collision operator options, including a simple drag-diffusion operatormore » in the parallel velocity space, Lorentz collisions, and a linearized model Fokker-Planck collision operator conserving momentum and energy (© 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)« less

Measurement and simulation of passive fast-ion D-alpha emission from the DIII-D tokamak

DOE PAGES

Bolte, Nathan G.; Heidbrink, William W.; Pace, David; ...

2016-09-14

Spectra of passive fast-ion D-alpha (FIDA) light from beam ions that charge exchange with background neutrals are measured and simulated. The fast ions come from three sources: ions that pass through the diagnostic sightlines on their first full orbit, an axisymmetric confined population, and ions that are expelled into the edge region by instabilities. A passive FIDA simulation (P-FIDASIM) is developed as a forward model for the spectra of the first-orbit fast ions and consists of an experimentally-validated beam deposition model, an ion orbit-following code, a collisional-radiative model, and a synthetic spectrometer. Model validation consists of the simulation of 86more » experimental spectra that are obtained using 6 different neutral beam fast-ion sources and 13 different lines of sight. Calibrated spectra are used to estimate the neutral density throughout the cross-section of the tokamak. The resulting 2D neutral density shows the expected increase toward each X-point with average neutral densities of 8 X 10 9 cm -3 at the plasma boundary and 1 X 10 11 cm -3 near the wall. Here, fast ions that are on passing orbits are expelled by the sawtooth instability more readily than trapped ions. In a sample discharge, approximately 1% of the fast-ion population is ejected into the high neutral density region per sawtooth crash.« less

Validation of conducting wall models using magnetic measurements

DOE PAGES

Hanson, Jeremy M.; Bialek, James M.; Turco, Francesca; ...

2016-08-16

The impact of conducting wall eddy currents on perturbed magnetic field measurements is a key issue for understanding the measurement and control of long-wavelength MHD stability in tokamak devices. As plasma response models have growth in sophistication, the need to understand and resolve small changes in these measurements has become more important, motivating increased fidelity in simulations of externally applied fields and the wall eddy current response. In this manuscript, we describe thorough validation studies of the wall models in the MARS-F and VALEN stability codes, using coil–sensor vacuum coupling measurements from the DIII-D tokamak. The valen formulation treats conductingmore » structures with arbitrary threedimensional geometries, while mars-f uses an axisymmetric wall model and a spectral decomposition of the problem geometry with a fixed toroidal harmonic n. The vacuum coupling measurements have a strong sensitivity to wall eddy currents induced by timechanging coil currents, owing to the close proximities of both the sensors and coils to the wall. Measurements from individual coil and sensor channels are directly compared with valen predictions. It is found that straightforward improvements to the valen model, such as refining the wall mesh and simulating the vertical extent of the DIII-D poloidal field sensors, lead to good agreement with the experimental measurements. In addition, couplings to multi-coil, n = 1 toroidal mode perturbations are calculated from the measurements and compared with predictions from both codes. Lastly, the toroidal mode comparisons favor the fully three-dimensional simulation approach, likely because this approach naturally treats n > 1 sidebands generated by the coils and wall eddy currents, as well as the n = 1 fundamental.« less

Validation of conducting wall models using magnetic measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanson, Jeremy M.; Bialek, James M.; Turco, Francesca

The impact of conducting wall eddy currents on perturbed magnetic field measurements is a key issue for understanding the measurement and control of long-wavelength MHD stability in tokamak devices. As plasma response models have growth in sophistication, the need to understand and resolve small changes in these measurements has become more important, motivating increased fidelity in simulations of externally applied fields and the wall eddy current response. In this manuscript, we describe thorough validation studies of the wall models in the MARS-F and VALEN stability codes, using coil–sensor vacuum coupling measurements from the DIII-D tokamak. The valen formulation treats conductingmore » structures with arbitrary threedimensional geometries, while mars-f uses an axisymmetric wall model and a spectral decomposition of the problem geometry with a fixed toroidal harmonic n. The vacuum coupling measurements have a strong sensitivity to wall eddy currents induced by timechanging coil currents, owing to the close proximities of both the sensors and coils to the wall. Measurements from individual coil and sensor channels are directly compared with valen predictions. It is found that straightforward improvements to the valen model, such as refining the wall mesh and simulating the vertical extent of the DIII-D poloidal field sensors, lead to good agreement with the experimental measurements. In addition, couplings to multi-coil, n = 1 toroidal mode perturbations are calculated from the measurements and compared with predictions from both codes. Lastly, the toroidal mode comparisons favor the fully three-dimensional simulation approach, likely because this approach naturally treats n > 1 sidebands generated by the coils and wall eddy currents, as well as the n = 1 fundamental.« less

A Monte-Carlo Benchmark of TRIPOLI-4® and MCNP on ITER neutronics

NASA Astrophysics Data System (ADS)

Blanchet, David; Pénéliau, Yannick; Eschbach, Romain; Fontaine, Bruno; Cantone, Bruno; Ferlet, Marc; Gauthier, Eric; Guillon, Christophe; Letellier, Laurent; Proust, Maxime; Mota, Fernando; Palermo, Iole; Rios, Luis; Guern, Frédéric Le; Kocan, Martin; Reichle, Roger

2017-09-01

Radiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the `C-lite', is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results.

A fast low-to-high confinement mode bifurcation dynamics in the boundary-plasma gyrokinetic code XGC1

DOE PAGES

Ku, S.; Chang, C. S.; Hager, R.; ...

2018-04-18

Here, a fast edge turbulence suppression event has been simulated in the electrostatic version of the gyrokinetic particle-in-cell code XGC1 in a realistic diverted tokamak edge geometry under neutral particle recycling. The results show that the sequence of turbulent Reynolds stress followed by neoclassical ion orbit-loss driven together conspire to form the sustaining radial electric field shear and to quench turbulent transport just inside the last closed magnetic flux surface. As a result, the main suppression action is located in a thin radial layer around ψ N≃0.96–0.98, where ψ N is the normalized poloidal flux, with the time scale ~0.1more » ms.« less

Suppression criteria of parasitic mode oscillations in a gyrotron beam tunnel

NASA Astrophysics Data System (ADS)

Kumar, Nitin; Singh, Udaybir; Singh, T. P.; Sinha, A. K.

2011-02-01

This paper presents the design criteria of the parasitic mode oscillations suppression for a periodic, ceramic, and copper loaded gyrotron beam tunnel. In such a type of beam tunnel, the suppression of parasitic mode oscillations is an important design problem. A method of beam-wave coupling coefficient and its mathematical formulation are presented. The developed design criteria are used in the beam tunnel design of a 42 GHz gyrotron to be developed for the Indian TOKAMAK system. The role of the thickness and the radius of the beam tunnel copper rings to obtain the developed design criteria are also discussed. The commercially available electromagnetic code CST and the electron trajectory code EGUN are used for the simulations.

Gyrokinetic micro-turbulence simulations on the NERSC 16-way SMP IBM SP computer: experiences and performance results

NASA Astrophysics Data System (ADS)

Ethier, Stephane; Lin, Zhihong

2001-10-01

Earlier this year, the National Energy Research Scientific Computing center (NERSC) took delivery of the second most powerful computer in the world. With its 2,528 processors running at a peak performance of 1.5 GFlops, this IBM SP machine has a theoretical performance of almost 3.8 TFlops. To efficiently harness such computing power in one single code is not an easy task and requires a good knowledge of the computer's architecture. Here we present the steps that we followed to improve our gyrokinetic micro-turbulence code GTC in order to take advantage of the new 16-way shared memory nodes of the NERSC IBM SP. Performance results are shown as well as details about the improved mixed-mode MPI-OpenMP model that we use. The enhancements to the code allowed us to tackle much bigger problem sizes, getting closer to our goal of simulating an ITER-size tokamak with both kinetic ions and electrons.(This work is supported by DOE Contract No. DE-AC02-76CH03073 (PPPL), and in part by the DOE Fusion SciDAC Project.)

Optimal Control Techniques for ResistiveWall Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Clement, Mitchell Dobbs Pearson

Tokamaks can excite kink modes that can lock or nearly lock to the vacuum vessel wall, and whose rotation frequencies and growth rates vary in time but are generally inversely proportional to the magnetic flux diffusion time of the vacuum vessel wall. This magnetohydrodynamic (MHD) instability is pressure limiting in tokamaks and is called the Resistive Wall Mode (RWM). Future tokamaks that are expected to operate as fusion reactors will be required to maximize plasma pressure in order to maximize fusion performance. The DIII-D tokamak is equipped with electromagnetic control coils, both inside and outside of its vacuum vessel, which create magnetic fields that are small by comparison to the machine's equilibrium field but are able to dynamically counteract the RWM. Presently for RWM feedback, DIII-D uses its interior control coils using a classical proportional gain only controller to achieve high plasma pressure. Future advanced tokamak designs will not likely have the luxury of interior control coils and a proportional gain algorithm is not expected to be effective with external control coils. The computer code VALEN was designed to calculate the performance of an MHD feedback control system in an arbitrary geometry. VALEN models the perturbed magnetic field from a single MHD instability and its interaction with surrounding conducting structures using a finite element approach. A linear quadratic gaussian (LQG) control, or H 2 optimal control, algorithm based on the VALEN model for RWM feedback was developed for use with DIII-D's external control coil set. The algorithm is implemented on a platform that combines a graphics processing unit (GPU) for real-time control computation with low latency digital input/output control hardware and operates in parallel with the DIII-D Plasma Control System (PCS). Simulations and experiments showed that modern control techniques performed better, using 77% less current, than classical techniques when using coils external to the vacuum vessel for RWM feedback. RWM feedback based on VALEN outperformed a classical control algorithm using external coils to suppress the normalized plasma response to a rotating n=1 perturbation applied by internal coils over a range of frequencies. This study describes the design, development and testing of the GPU based control hardware and algorithm along with its performance during experiment and simulation.

ALCBEAM - Neutral beam formation and propagation code for beam-based plasma diagnostics

NASA Astrophysics Data System (ADS)

Bespamyatnov, I. O.; Rowan, W. L.; Liao, K. T.

2012-03-01

ALCBEAM is a new three-dimensional neutral beam formation and propagation code. It was developed to support the beam-based diagnostics installed on the Alcator C-Mod tokamak. The purpose of the code is to provide reliable estimates of the local beam equilibrium parameters: such as beam energy fractions, density profiles and excitation populations. The code effectively unifies the ion beam formation, extraction and neutralization processes with beam attenuation and excitation in plasma and neutral gas and beam stopping by the beam apertures. This paper describes the physical processes interpreted and utilized by the code, along with exploited computational methods. The description is concluded by an example simulation of beam penetration into plasma of Alcator C-Mod. The code is successfully being used in Alcator C-Mod tokamak and expected to be valuable in the support of beam-based diagnostics in most other tokamak environments. Program summaryProgram title: ALCBEAM Catalogue identifier: AEKU_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEKU_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 66 459 No. of bytes in distributed program, including test data, etc.: 7 841 051 Distribution format: tar.gz Programming language: IDL Computer: Workstation, PC Operating system: Linux RAM: 1 GB Classification: 19.2 Nature of problem: Neutral beams are commonly used to heat and/or diagnose high-temperature magnetically-confined laboratory plasmas. An accurate neutral beam characterization is required for beam-based measurements of plasma properties. Beam parameters such as density distribution, energy composition, and atomic excited populations of the beam atoms need to be known. Solution method: A neutral beam is initially formed as an ion beam which is extracted from the ion source by high voltage applied to the extraction and accelerating grids. The current distribution of a single beamlet emitted from a single pore of IOS depends on the shape of the plasma boundary in the emission region. Total beam extracted by IOS is calculated at every point of 3D mesh as sum of all contributions from each grid pore. The code effectively unifies the ion beam formation, extraction and neutralization processes with neutral beam attenuation and excitation in plasma and neutral gas and beam stopping by the beam apertures. Running time: 10 min for a standard run.

Measurements of localized core turbulence & turbulence suppression on DIII-D

NASA Astrophysics Data System (ADS)

Shafer, Morgan W.

The crucial dynamics of turbulent-driven cross-field transport in tokamak plasmas reside in the two-dimensional (2D) radial/poloidal plane. Thus, 2D measurements of turbulence are needed to test theoretical models and validate sophisticated gyrokinetic codes. Furthermore, measurements are important for understanding the role of turbulence suppression in enhanced confinement regimes. The Beam Emission Spectroscopy (BES) diagnostic on the DIII-D tokamak measures localized, long-wavelength (k⊥rho i≤1) density fluctuations in the 2D radial/poloidal plane and is suitable for these studies. Measurements of turbulence amplitude, S(kr,k theta) spectra, correlation lengths, decorrelation rates and group velocities are obtained via BES in the core (0.3< r/a <0.9) and compared to nonlinear gyrokinetic simulations from the GYRO code. The 2D measurements show a tilted eddy structure in the core that is consistent with ExB shear. The S(kr,ktheta) spectra are directly compared to GYRO simulations. These comparisons show the 2D structure is in reasonable agreement at r/a = 0.5 where the predicted turbulence amplitude and heat flux agree well with the measurements. However, the simulations show a strongly tilted eddy structure that extends to high-kr at r/a = 0.75, where the simulations under-predict the turbulence amplitude and heat flux. This is not observed in the experiment and suggests a possible over-exaggeration of an ExB or zonal flow shearing mechanism in the simulations. Measurements demonstrate local turbulence suppression near low-order rational q-surfaces at low magnetic shear. This interaction can lead to an Internal Transport Barrier (ITB) provided sufficient equilibrium ExB shear (largely due to the toroidal rotation of neutral beam heated rotating plasmas) sustains the barrier. Related GYRO simulations suggest these ITBs are triggered by zonal flows that form near the q = 2 surface. Consistent with the simulations, localized measurements demonstrate increased shear in the poloidal turbulence velocity. The resulting shear rate transiently exceeds the decorrelation rate, causing a reduction in turbulence and radial correlation length. The layer of suppressed turbulence moves radially outward, nearly coincident with integer q-surfaces.

Magnetic flux pumping in 3D nonlinear magnetohydrodynamic simulations

NASA Astrophysics Data System (ADS)

Krebs, I.; Jardin, S. C.; Günter, S.; Lackner, K.; Hoelzl, M.; Strumberger, E.; Ferraro, N.

2017-10-01

A self-regulating magnetic flux pumping mechanism in tokamaks that maintains the core safety factor at q ≈1 , thus preventing sawteeth, is analyzed in nonlinear 3D magnetohydrodynamic simulations using the M3D-C1 code. In these simulations, the most important mechanism responsible for the flux pumping is that a saturated (m =1 ,n =1 ) quasi-interchange instability generates an effective negative loop voltage in the plasma center via a dynamo effect. It is shown that sawtoothing is prevented in the simulations if β is sufficiently high to provide the necessary drive for the (m =1 ,n =1 ) instability that generates the dynamo loop voltage. The necessary amount of dynamo loop voltage is determined by the tendency of the current density profile to centrally peak which, in our simulations, is controlled by the peakedness of the applied heat source profile.

Magnetic flux pumping in 3D nonlinear magnetohydrodynamic simulations

DOE PAGES

Krebs, I.; Jardin, S. C.; Gunter, S.; ...

2017-09-27

A self-regulating magnetic flux pumping mechanism in tokamaks that maintains the core safety factor at q≈1, thus preventing sawteeth, is analyzed in nonlinear 3D magnetohydrodynamic simulations using the M3D-C1 code. In these simulations, the most important mechanism responsible for the flux pumping is that a saturated (m=1,n=1) quasi-interchange instability generates an effective negative loop voltage in the plasma center via a dynamo effect. It is shown that sawtoothing is prevented in the simulations if β is sufficiently high to provide the necessary drive for the (m=1,n=1) instability that generates the dynamo loop voltage. In conclusion, the necessary amount of dynamomore » loop voltage is determined by the tendency of the current density profile to centrally peak which, in our simulations, is controlled by the peakedness of the applied heat source profile.« less

Nonlinear Full-f Edge Gyrokinetic Turbulence Simulations

NASA Astrophysics Data System (ADS)

Xu, X. Q.; Dimits, A. M.; Umansky, M. V.

2008-11-01

TEMPEST is a nonlinear full-f 5D electrostatic gyrokinetic code for simulations of neoclassical and turbulent transport for tokamak plasmas. Given an initial density perturbation, 4D TEMPEST simulations show that the kinetic GAM exists in the edge in the form of outgoing waves [1], its radial scale is set by plasma profiles, and the ion temperature inhomogeneity is necessary for GAM radial propagation. From an initial Maxwellian distribution with uniform poloidal profiles on flux surfaces, the 5D TEMPEST simulations in a flux coordinates with Boltzmann electron model in a circular geometry show the development of neoclassical equilibrium, the generation of the neoclassical electric field due to neoclassical polarization, and followed by a growth of instability due to the spatial gradients. 5D TEMPEST simulations of kinetic GAM turbulent generation, radial propagation, and its impact on transport will be reported. [1] X. Q. Xu, Phys. Rev. E., 78 (2008).

Anomalous transport scaling in the DIII-D tokamak matched by supercomputer simulation.

PubMed

Candy, J; Waltz, R E

2003-07-25

Gyrokinetic simulation of tokamak transport has evolved sufficiently to allow direct comparison of numerical results with experimental data. It is to be emphasized that only with the simultaneous inclusion of many distinct and complex effects can this comparison realistically be made. Until now, numerical studies of tokamak microturbulence have been restricted to either (a) flux tubes or (b) electrostatic fluctuations. Using a newly developed global electromagnetic solver, we have been able to recover via direct simulation the Bohm-like scaling observed in DIII-D L-mode discharges. We also match, well within experimental uncertainty, the measured energy diffusivities.

Computation of Alfvèn eigenmode stability and saturation through a reduced fast ion transport model in the TRANSP tokamak transport code

NASA Astrophysics Data System (ADS)

Podestà, M.; Gorelenkova, M.; Gorelenkov, N. N.; White, R. B.

2017-09-01

Alfvénic instabilities (AEs) are well known as a potential cause of enhanced fast ion transport in fusion devices. Given a specific plasma scenario, quantitative predictions of (i) expected unstable AE spectrum and (ii) resulting fast ion transport are required to prevent or mitigate the AE-induced degradation in fusion performance. Reduced models are becoming an attractive tool to analyze existing scenarios as well as for scenario prediction in time-dependent simulations. In this work, a neutral beam heated NSTX discharge is used as reference to illustrate the potential of a reduced fast ion transport model, known as kick model, that has been recently implemented for interpretive and predictive analysis within the framework of the time-dependent tokamak transport code TRANSP. Predictive capabilities for AE stability and saturation amplitude are first assessed, based on given thermal plasma profiles only. Predictions are then compared to experimental results, and the interpretive capabilities of the model further discussed. Overall, the reduced model captures the main properties of the instabilities and associated effects on the fast ion population. Additional information from the actual experiment enables further tuning of the model’s parameters to achieve a close match with measurements.

Computation of Alfvèn eigenmode stability and saturation through a reduced fast ion transport model in the TRANSP tokamak transport code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Podestà, M.; Gorelenkova, M.; Gorelenkov, N. N.

Alfvénic instabilities (AEs) are well known as a potential cause of enhanced fast ion transport in fusion devices. Given a specific plasma scenario, quantitative predictions of (i) expected unstable AE spectrum and (ii) resulting fast ion transport are required to prevent or mitigate the AE-induced degradation in fusion performance. Reduced models are becoming an attractive tool to analyze existing scenarios as well as for scenario prediction in time-dependent simulations. Here, in this work, a neutral beam heated NSTX discharge is used as reference to illustrate the potential of a reduced fast ion transport model, known as kick model, that hasmore » been recently implemented for interpretive and predictive analysis within the framework of the time-dependent tokamak transport code TRANSP. Predictive capabilities for AE stability and saturation amplitude are first assessed, based on given thermal plasma profiles only. Predictions are then compared to experimental results, and the interpretive capabilities of the model further discussed. Overall, the reduced model captures the main properties of the instabilities and associated effects on the fast ion population. Finally, additional information from the actual experiment enables further tuning of the model's parameters to achieve a close match with measurements.« less

Computation of Alfvèn eigenmode stability and saturation through a reduced fast ion transport model in the TRANSP tokamak transport code

DOE PAGES

Podestà, M.; Gorelenkova, M.; Gorelenkov, N. N.; ...

2017-07-20

Alfvénic instabilities (AEs) are well known as a potential cause of enhanced fast ion transport in fusion devices. Given a specific plasma scenario, quantitative predictions of (i) expected unstable AE spectrum and (ii) resulting fast ion transport are required to prevent or mitigate the AE-induced degradation in fusion performance. Reduced models are becoming an attractive tool to analyze existing scenarios as well as for scenario prediction in time-dependent simulations. Here, in this work, a neutral beam heated NSTX discharge is used as reference to illustrate the potential of a reduced fast ion transport model, known as kick model, that hasmore » been recently implemented for interpretive and predictive analysis within the framework of the time-dependent tokamak transport code TRANSP. Predictive capabilities for AE stability and saturation amplitude are first assessed, based on given thermal plasma profiles only. Predictions are then compared to experimental results, and the interpretive capabilities of the model further discussed. Overall, the reduced model captures the main properties of the instabilities and associated effects on the fast ion population. Finally, additional information from the actual experiment enables further tuning of the model's parameters to achieve a close match with measurements.« less

Modeling ECCD/MHD coupling using NIMROD, GENRAY, and the Integrated Plasma Simulator

NASA Astrophysics Data System (ADS)

Jenkins, Thomas G.; Schnack, D. D.; Sovinec, C. R.; Hegna, C. C.; Callen, J. D.; Ebrahimi, F.; Kruger, S. E.; Carlsson, J.; Held, E. D.; Ji, J.-Y.; Harvey, R. W.; Smirnov, A. P.; Elwasif, W. R.

2009-11-01

We summarize ongoing theoretical/numerical work relevant to the development of a self--consistent framework for the inclusion of RF effects in fluid simulations; specifically, we consider the stabilization of resistive tearing modes in tokamak geometry by electron cyclotron current drive. In the fluid equations, ad hoc models for the RF--induced currents have previously been shown to shrink or altogether suppress the nonlinearly saturated magnetic islands generated by tearing modes; progress toward a self--consistent model is reported. The interfacing of the NIMROD [1] code with the GENRAY/CQL3D [2] codes (which calculate RF propagation and energy/momentum deposition) via the Integrated Plasma Simulator (IPS) framework [3] is explained, RF-induced rational surface motion and the equilibration of RF--induced currents over plasma flux surfaces are investigated, and the efficient reduction of saturated island widths through time modulation and spatial localization of the ECCD is explored. [1] Sovinec et al., JCP 195, 355 (2004) [2]www.compxco.com [3] Both the IPS development and the research presented here are part of the SWIM project. Funded by U.S. DoE.

Modeling of RF/MHD coupling using NIMROD, GENRAY, and the Integrated Plasma Simulator

NASA Astrophysics Data System (ADS)

Jenkins, Thomas; Schnack, D. D.; Sovinec, C. R.; Hegna, C. C.; Callen, J. D.; Ebrahimi, F.; Kruger, S. E.; Carlsson, J.; Held, E. D.; Ji, J.-Y.; Harvey, R. W.; Smirnov, A. P.

2009-05-01

We summarize ongoing theoretical/numerical work relevant to the development of a self--consistent framework for the inclusion of RF effects in fluid simulations; specifically considering resistive tearing mode stabilization in tokamak (DIII--D--like) geometry via ECCD. Relatively simple (though non--self--consistent) models for the RF--induced currents are incorporated into the fluid equations, markedly reducing the width of the nonlinearly saturated magnetic islands generated by tearing modes. We report our progress toward the self--consistent modeling of these RF--induced currents. The initial interfacing of the NIMROD* code with the GENRAY/CQL3D** codes (calculating RF propagation and energy/momentum deposition) via the Integrated Plasma Simulator (IPS) framework*** is explained, equilibration of RF--induced currents over the plasma flux surfaces is investigated, and studies exploring the efficient reduction of saturated island widths through time modulation and spatial localization of the ECCD are presented. *[Sovinec et al., JCP 195, 355 (2004)] **[www.compxco.com] ***[This research and the IPS development are both part of the SWIM project. Funded by U.S. DoE.

Wall-touching kink mode calculations with the M3D code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Breslau, J. A., E-mail: jbreslau@pppl.gov; Bhattacharjee, A.

This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicatedmore » by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.« less

High frequency RF waves

NASA Astrophysics Data System (ADS)

Horton, William; Brookman, M.; Goniche, M.; Peysson, Y.; Ekedahl, A.

2017-10-01

ECH and LHCD- are scattered by the density and magnetic field turbulence from drift waves as measured in and Tore Supra-WEST, EAST and DIII-D. Ray equations give the spreading from plasma refraction from the antenna through the core plasma until and change the parallel phase velocity evolves to where RF waves are absorbed by the electrons. Extensive LH ray tracing and absorption has been reported using the coupled CP3O ray tracing and LUKE electron phase space density code with collisionless electron-wave resonant absorption. In theory and simulations are shown for the ray propagation with the resulting electron distributions along with the predicted X ray distribution that compared to the measured X-ray spectrum. Lower-hybrid is essential for steady-state operation in tokamaks with control of the high-energy electrons intrinsic to tokamaks confinement and heating. The record steady tokamak plasma is Tore Supra a steady 6 minute steady state plasma with 1 Gigajoule energy passing through the plasma. WEST is repeating the experiments with ITER shaped separatrix and divertor chamber and EAST achieved comparable long-pulse plasmas. Results are presented from an IFS-3D spectral code with a pair of inside-outside LHCD antennas and a figure-8 magnetic separatrix are presented. Scattering of the slow wave into the fast wave wave is explored showing the RF scattering from drift wave dne and dB increases the core penetration may account the measured broad X-ray spectrum. Work supported by the DoE through Grants to the Institute for Fusion Studies [DE-FG02-04ER54742], ARLUT and General Atomics, San Diego, California, USA and the IRFM at Cadarache by the Comissariat Energie Atomique, France.

Sawtooth mitigation in 3D MHD tokamak modelling with applied magnetic perturbations

NASA Astrophysics Data System (ADS)

Bonfiglio, D.; Veranda, M.; Cappello, S.; Chacón, L.; Escande, D. F.

2017-01-01

The effect of magnetic perturbations (MPs) on the sawtoothing dynamics of the internal kink mode in the tokamak is discussed in the framework of nonlinear 3D MHD modelling. Numerical simulations are performed with the pixie3d code (Chacón 2008 Phys. Plasmas 15 056103) based on a D-shaped configuration in toroidal geometry. MPs are applied as produced by two sets of coils distributed along the toroidal direction, one set located above and the other set below the outboard midplane, like in experimental devices such as DIII-D and ASDEX Upgrade. The capability of n  =  1 MPs to affect quasi-periodic sawteeth is shown to depend on the toroidal phase difference Δ φ between the perturbations produced by the two sets of coils. In particular, sawtooth mitigation is obtained for the Δ φ =π phasing, whereas no significant effect is observed for Δ φ =0 . Numerical findings are explained by the interplay between different poloidal harmonics in the spectrum of applied MPs, and appear to be consistent with experiments performed in the DIII-D device. Sawtooth mitigation and stimulation of self-organized helical states by applied MPs have been previously demonstrated in both circular tokamak and reversed-field pinch (RFP) experiments in the RFX-mod device, and in related 3D MHD modelling.

Dynamics of kinetic geodesic-acoustic modes and the radial electric field in tokamak neoclassical plasmas

NASA Astrophysics Data System (ADS)

Xu, X. Q.; Belli, E.; Bodi, K.; Candy, J.; Chang, C. S.; Cohen, R. H.; Colella, P.; Dimits, A. M.; Dorr, M. R.; Gao, Z.; Hittinger, J. A.; Ko, S.; Krasheninnikov, S.; McKee, G. R.; Nevins, W. M.; Rognlien, T. D.; Snyder, P. B.; Suh, J.; Umansky, M. V.

2009-06-01

We present edge gyrokinetic simulations of tokamak plasmas using the fully non-linear (full-f) continuum code TEMPEST. A non-linear Boltzmann model is used for the electrons. The electric field is obtained by solving the 2D gyrokinetic Poisson equation. We demonstrate the following. (1) High harmonic resonances (n > 2) significantly enhance geodesic-acoustic mode (GAM) damping at high q (tokamak safety factor), and are necessary to explain the damping observed in our TEMPEST q-scans and consistent with the experimental measurements of the scaling of the GAM amplitude with edge q95 in the absence of obvious evidence that there is a strong q-dependence of the turbulent drive and damping of the GAM. (2) The kinetic GAM exists in the edge for steep density and temperature gradients in the form of outgoing waves, its radial scale is set by the ion temperature profile, and ion temperature inhomogeneity is necessary for GAM radial propagation. (3) The development of the neoclassical electric field evolves through different phases of relaxation, including GAMs, their radial propagation and their long-time collisional decay. (4) Natural consequences of orbits in the pedestal and scrape-off layer region in divertor geometry are substantial non-Maxwellian ion distributions and parallel flow characteristics qualitatively like those observed in experiments.

Simulations of drift-Alfven turbulence in LAPD using BOUT

NASA Astrophysics Data System (ADS)

Popovich, Pavel; Umansky, Maxim; Carter, Troy; Cowley, Steve

2008-11-01

The LArge Plasma Device (LAPD) at UCLA is a 17 m long, 60 cm diameter magnetized plasma column with typical plasma parameters ne˜1x10^12cm-3, Te˜10eV, and B ˜1kG. The simple geometry and extensive measurement capability on LAPD allows for detailed comparison with and validation of numerical simulations of turbulence and transport. We analyse the LAPD results using simulations with the boundary plasma turbulence code BOUT. BOUT models the 3D electromagnetic plasma turbulence solving a system of fluid moment equations in a general tokamak geometry near the boundary. We will discuss the physical model and modifications of the BOUT code required for the LAPD configuration, and present the first results of the simulations and comparison to experimental measurements. In particular, a confinement transition is observed in LAPD under the application of bias-driven rotation. Also, intermittent generation and convection of filamentary structures (``blobs'' and ``holes'') is observed in the LAPD edge. Application of BOUT to modeling of these two phenomena will be discussed. E. Maggs, T.A. Carter, and R.J. Taylor, Phys. Plasmas 14, (2007) T.A. Carter, Phys. Plasmas 13, (2006)

Bifurcated helical core equilibrium states in tokamaks

NASA Astrophysics Data System (ADS)

Cooper, W. A.; Chapman, I. T.; Schmitz, O.; Turnbull, A. D.; Tobias, B. J.; Lazarus, E. A.; Turco, F.; Lanctot, M. J.; Evans, T. E.; Graves, J. P.; Brunetti, D.; Pfefferlé, D.; Reimerdes, H.; Sauter, O.; Halpern, F. D.; Tran, T. M.; Coda, S.; Duval, B. P.; Labit, B.; Pochelon, A.; Turnyanskiy, M. R.; Lao, L.; Luce, T. C.; Buttery, R.; Ferron, J. R.; Hollmann, E. M.; Petty, C. C.; van Zeeland, M.; Fenstermacher, M. E.; Hanson, J. M.; Lütjens, H.

2013-07-01

Tokamaks with weak to moderate reversed central shear in which the minimum inverse rotational transform (safety factor) qmin is in the neighbourhood of unity can trigger bifurcated magnetohydrodynamic equilibrium states, one of which is similar to a saturated ideal internal kink mode. Peaked prescribed pressure profiles reproduce the ‘snake’ structures observed in many tokamaks which has led to a novel explanation of the snake as a bifurcated equilibrium state. Snake equilibrium structures are computed in simulations of the tokamak à configuration variable (TCV), DIII-D and mega amp spherical torus (MAST) tokamaks. The internal helical deformations only weakly modulate the plasma-vacuum interface which is more sensitive to ripple and resonant magnetic perturbations. On the other hand, the external perturbations do not alter the helical core deformation in a significant manner. The confinement of fast particles in MAST simulations deteriorate with the amplitude of the helical core distortion. These three-dimensional bifurcated solutions constitute a paradigm shift that motivates the applications of tools developed for stellarator research in tokamak physics investigations.

Local and integral disruption forces on the tokamak wall

NASA Astrophysics Data System (ADS)

Pustovitov, V. D.; Kiramov, D. I.

2018-04-01

The disruption-induced forces on the tokamak wall are evaluated analytically within the standard large-aspect-ratio model that implies axisymmetry, circular plasma and wall, and absence of halo currents. Additionally, the ideal-wall reaction is assumed. The disruptions are modelled as rapid changes in the plasma pressure (thermal quench (TQ)) and net current (current quench (CQ)). The force distribution over the poloidal angle is found as a function of these inputs. The derived formulas allow comparison of the TQ- and CQ-produced forces calculated differently, with and without account of the poloidal current induced in the wall. The latter variant represents the inherent property of the codes treating the wall as a set of toroidal filaments. It is proved here that such a simplification leads to unacceptably large errors in the simulated forces for both TQs and CQs. It is also shown that the TQ part of the force must prevail over that due to CQ in the high-β scenarios developed for JT-60SA and ITER.

Plasma density injection and flow during coaxial helicity injection in a tokamak

NASA Astrophysics Data System (ADS)

Hooper, E. B.

2018-02-01

Whole device, resistive MHD simulations of spheromaks and tokamaks have used a large diffusion coefficient that maintains a nearly constant density throughout the device. In the present work, helicity and plasma are coinjected into a low-density plasma in a tokamak with a small diffusion coefficient. As in previous simulations [Hooper et al., Phys. Plasmas 20, 092510 (2013)], a flux bubble is formed, which expands to fill the tokamak volume. The injected plasma is non-uniform inside the bubble. The flow pattern is analyzed; when the simulation is not axisymmetric, an n = 1 mode on the surface of the bubble generates leakage of plasma into the low-density volume. Closed flux is generated following injection, as in experiments and previous simulations. The result provides a more detailed physics analysis of the injection, including density non-uniformities in the plasma that may affect its use as a startup plasma [Raman et al., Phys. Rev. Lett. 97, 175002 (2006)].

Study of Globus-M Tokamak Poloidal System and Plasma Position Control

NASA Astrophysics Data System (ADS)

Dokuka, V. N.; Korenev, P. S.; Mitrishkin, Yu. V.; Pavlova, E. A.; Patrov, M. I.; Khayrutdinov, R. R.

2017-12-01

In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

Molecular Dynamics Simulation of Hydrogen Trapping on Sigma 5 Tungsten Grain Boundaries

NASA Astrophysics Data System (ADS)

Al-Shalash, Aws Mohammed Taha

Tungsten as a plasma facing material is the predominant contender for future Tokamak reactor environments. The interaction between the plasma particles and tungsten is crucial to be studied for successful usage and design of tungsten in the plasma facing components ensuring the reliability and longevity of the fusion reactors. The bombardment of the sigma 5 polycrystalline tungsten was modeled using the molecular dynamics simulation through the large-scale atomic/molecular massively parallel simulator (LAMMPS) code and Tersoff type interatomic potential. By simulating the operational conditions of the Tokamak reactors, the hydrogen trapping rate, implantation distribution, and bubble formation was investigated at various temperatures (300-1200 K) and various hydrogen incident energy (20-100 eV). The substrate's temperature increases the deflected H atoms, and increases the penetration depth for the ones that go through. As well, the lower temperature tungsten substrates retain more H atoms. Increasing the bombarded hydrogen's energy increases the trapping and retention rate and the depth of penetration. Another experiments were conducted to determine whether the Sigma5 grain boundary's (GB) location affects the trapping profiles in H. The findings are ranges from small effect on deflection rates at low H energies to no effect at high H energies. However, there is a considerable effect on shifting the trapping depth profile upward toward the surface when raising the GB closer to the surface. Hydrogen atoms are highly mobile on tungsten substrate, yet no bubble formation was witnessed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

H.E. Mynick, N. Pomphrey and P. Xanthopoulos

Recent progress in reducing turbulent transport in stellarators and tokamaks by 3D shaping using a stellarator optimization code in conjunction with a gyrokinetic code is presented. The original applications of the method focussed on ion temperature gradient transport in a quasi-axisymmetric stellarator design. Here, an examination of both other turbulence channels and other starting configurations is initiated. It is found that the designs evolved for transport from ion temperature gradient turbulence also display reduced transport from other transport channels whose modes are also stabilized by improved curvature, such as electron temperature gradient and ballooning modes. The optimizer is also appliedmore » to evolving from a tokamak, finding appreciable turbulence reduction for these devices as well. From these studies, improved understanding is obtained of why the deformations found by the optimizer are beneficial, and these deformations are related to earlier theoretical work in both stellarators and tokamaks.« less

Ideal MHD Stability Prediction and Required Power for EAST Advanced Scenario

NASA Astrophysics Data System (ADS)

Chen, Junjie; Li, Guoqiang; Qian, Jinping; Liu, Zixi

2012-11-01

The Experimental Advanced Superconducting Tokamak (EAST) is the first fully superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. The ideal magnetohydrodynamic (MHD) stability and required power for the EAST advanced tokamak (AT) scenario with negative central shear and double transport barrier (DTB) are investigated. With the equilibrium code TOQ and stability code GATO, the ideal MHD stability is analyzed. It is shown that a moderate ratio of edge transport barriers' (ETB) height to internal transport barriers' (ITBs) height is beneficial to ideal MHD stability. The normalized beta βN limit is about 2.20 (without wall) and 3.70 (with ideal wall). With the scaling law of energy confinement time, the required heating power for EAST AT scenario is calculated. The total heating power Pt increases as the toroidal magnetic field BT or the normalized beta βN is increased.

Final technical report for DE-SC00012633 AToM (Advanced Tokamak Modeling)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holland, Christopher; Orlov, Dmitri; Izzo, Valerie

This final report for the AToM project documents contributions from University of California, San Diego researchers over the period of 9/1/2014 – 8/31/2017. The primary focus of these efforts was on performing validation studies of core tokamak transport models using the OMFIT framework, including development of OMFIT workflow scripts. Additional work was performed to develop tools for use of the nonlinear magnetohydrodynamics code NIMROD in OMFIT, and its use in the study of runaway electron dynamics in tokamak disruptions.

Aspect ratio effects on limited scrape-off layer plasma turbulence

NASA Astrophysics Data System (ADS)

Jolliet, Sébastien; Halpern, Federico D.; Loizu, Joaquim; Mosetto, Annamaria; Ricci, Paolo

2014-02-01

The drift-reduced Braginskii model describing turbulence in the tokamak scrape-off layer is written for a general magnetic configuration with a limiter. The equilibrium is then specified for a circular concentric magnetic geometry retaining aspect ratio effects. Simulations are then carried out with the help of the global, flux-driven fluid three-dimensional code GBS [Ricci et al., Plasma Phys. Controlled Fusion 54, 124047 (2012)]. Linearly, both simulations and simplified analytical models reveal a stabilization of ballooning modes. Nonlinearly, flux-driven nonlinear simulations give a pressure characteristic length whose trends are correctly captured by the gradient removal theory [Ricci and Rogers, Phys. Plasmas 20, 010702 (2013)], that assumes the profile flattening from the linear modes as the saturation mechanism. More specifically, the linear stabilization of ballooning modes is reflected by a 15% increase in the steady-state pressure gradient obtained from GBS nonlinear simulations when going from an infinite to a realistic aspect ratio.

GTC simulations of ion temperature gradient driven instabilities in W7-X and LHD stellarators

NASA Astrophysics Data System (ADS)

Wang, Hongyu

2017-10-01

We report GTC linear simulations of ion temperature gradient (ITG) instabilities in Wendelstein 7-X (W7-X) and Large Helical Device (LHD) stellarators. GTC has recently been updated to treat 3D equilibria by interfacing with MHD equilibrium code VMEC. GTC simulations of ITG have been carried out in both full torus and partial torus taking into account the toroidal periodicity of the stellarators. The effects of toroidal mode coupling on linear dispersions and mode structures in W7-X and LHD are studied. The mode structure in W7-X is more localized in the toroidal direction, and LHD is more extended in the toroidal direction and tokamak-like. Linear growth rates, real frequencies, and mode structures agree reasonably with results of EUTERPE simulations. In collaboration with I. Holod, J. Riemann, Z. Lin, J. Bao, L. Shi, S. Taimourzadeh, R. Kleiber, and M. Borchardt.

Simulation of ITG instabilities with fully kinetic ions and drift-kinetic electrons in tokamaks

NASA Astrophysics Data System (ADS)

Hu, Youjun; Chen, Yang; Parker, Scott

2017-10-01

A turbulence simulation model with fully kinetic ions and drift-kinetic electrons is being developed in the toroidal electromagnetic turbulence code GEM. This is motivated by the observation that gyrokinetic ions are not well justified in simulating turbulence in tokamak edges with steep density profile, where ρi / L is not small enough to be used a small parameter needed by the gyrokinetic ordering (here ρi is the gyro-radius of ions and L is the scale length of density profile). In this case, the fully kinetic ion model may be useful. Our model uses an implicit scheme to suppress high-frequency compressional Alfven waves and waves associated with the gyro-motion of ions. The ion orbits are advanced by using the well-known Boris scheme, which reproduces correct drift-motion even with large time-step comparable to the ion gyro-period. The field equation in this model is Ampere's law with the magnetic field eliminated by using an implicit scheme of Faraday's law. The current contributed by ions are computed by using an implicit δf method. A flux tube approximation is adopted, which makes the field equation much easier to solve. Numerical results of electromagnetic ITG obtained from this model will be presented and compared with the gyrokinetic results. This work is supported by U.S. Department of Energy, Office of Fusion Energy Sciences under Award No. DE-SC0008801.

Modelling ion cyclotron emission from KSTAR tokamak and LHD helical device plasmas

NASA Astrophysics Data System (ADS)

Dendy, Richard; Chapman, Ben; Reman, Bernard; Chapman, Sandra; Akiyama, Tsuyoshi; Yun, Gunsu

2017-10-01

New high quality measurements of ion cyclotron emission (ICE) from KSTAR and LHD greatly extend the scope and diversity of plasma conditions under which ICE is observed. Variables include the origin (fusion reactions or neutral beam injection) and energy (sub- or super-Alfvénic) of the minority energetic ions that drive ICE; the composition of the bulk plasma (hydrogen or deuterium) which supports the modes excited; plasma density in the emitting region, and the timescale on which it changes; and toroidal magnetic field geometry (tokamak or helical device). Future exploitation of ICE as a diagnostic for energetic ion populations in JET D-T plasmas and in ITER rests on quantitative understanding of the physics of the emission. This is tested and extended by current KSTAR and LHD measurements of ICE. We report progress on direct numerical simulation using full orbit ion kinetic codes that solve the Maxwell-Lorentz equations for hundreds of millions of particles. In the saturated regime, these simulations yield excited field spectra that correspond directly to the measured ICE spectra under diverse KSTAR and LHD regimes. At early times, comparison of simulation outputs with linear analytical theory confirms the magnetoacoustic cyclotron instability as the basic driver of ICE. Supported by RCUK Energy Programme Grant EP/P012450/1, NRF Korea Grant 2014M1A7A1A03029881, NIFS budget ULHH029 and Euratom.

Modelling of radiation impact on ITER Beryllium wall

NASA Astrophysics Data System (ADS)

Landman, I. S.; Janeschitz, G.

2009-04-01

In the ITER H-Mode confinement regime, edge localized instabilities (ELMs) will perturb the discharge. Plasma lost after each ELM moves along magnetic field lines and impacts on divertor armour, causing plasma contamination by back propagating eroded carbon or tungsten. These impurities produce enhanced radiation flux distributed mainly over the beryllium main chamber wall. The simulation of the complicated processes involved are subject of the integrated tokamak code TOKES that is currently under development. This work describes the new TOKES model for radiation transport through confined plasma. Equations for level populations of the multi-fluid plasma species and the propagation of different kinds of radiation (resonance, recombination and bremsstrahlung photons) are implemented. First simulation results without account of resonance lines are presented.

Center for the Study of Plasma Microturbulence

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parker, Scott E.

We have discovered a possible "natural fueling" mechanism in tokamak fusion reactors using large scale gyrokinetic turbulence simulation. In the presence of a heat flux dominated tokamak plasma, cold ions naturally pinch radially inward. If cold DT fuel is introduced near the edge using shallow pellet injection, the cold fuel will pinch inward, at the expense of hot helium ash going radially outward. By adjusting the cold DT fuel concentration, the core DT density profiles can be maintained. We have also shown that cold source ions from edge recycling of cold neutrals are pinched radially inward. This mechanism may bemore » important for fully understanding the edge pedestal buildup after an ELM crash. Work includes benchmarking the gyrokinetic turbulence codes in the electromagnetic regime. This includes cyclone base case parameters with an increasing plasma beta. The code comparisons include GEM, GYRO and GENE. There is good linear agreement between the codes using the Cyclone base case, but including electromagnetics and scanning the plasma beta. All the codes have difficulty achieving nonlinear saturation as the kinetic ballooning limit is approached. GEM does not saturate well when beta gets above about 1/2 of the ideal ballooning limit. We find that the lack of saturation is due to the long wavelength k{sub y} modes being nonlinearly pumped to high levels. If the fundamental k{sub y} mode is zeroed out, higher values of beta nonlinearly saturate well. Additionally, there have been studies to better understand CTEM nonlinear saturation and the importance of zonal flows. We have continued our investigation of trapped electron mode (TEM) turbulence. More recently, we have focused on the nonlinear saturation of TEM turbulence. An important feature of TEM is that in many parameter regimes, the zonal flow is unimportant. We find that when zonal flows are unimportant, zonal density is the dominant saturation mechanism. We developed a simple theory that agrees with the simulation and predicts zonal density generation and feedback stabilization of the most unstable mode even in the absence of zonal flow. We are using GEM to simulate NSTX discharges. We have also done verification and validation on DIII-D. Good agreement with GYRO and DIII-D flux levels were reported in the core region.« less

Nonlinear 3D visco-resistive MHD modeling of fusion plasmas: a comparison between numerical codes

NASA Astrophysics Data System (ADS)

Bonfiglio, D.; Chacon, L.; Cappello, S.

2008-11-01

Fluid plasma models (and, in particular, the MHD model) are extensively used in the theoretical description of laboratory and astrophysical plasmas. We present here a successful benchmark between two nonlinear, three-dimensional, compressible visco-resistive MHD codes. One is the fully implicit, finite volume code PIXIE3D [1,2], which is characterized by many attractive features, notably the generalized curvilinear formulation (which makes the code applicable to different geometries) and the possibility to include in the computation the energy transport equation and the extended MHD version of Ohm's law. In addition, the parallel version of the code features excellent scalability properties. Results from this code, obtained in cylindrical geometry, are compared with those produced by the semi-implicit cylindrical code SpeCyl, which uses finite differences radially, and spectral formulation in the other coordinates [3]. Both single and multi-mode simulations are benchmarked, regarding both reversed field pinch (RFP) and ohmic tokamak magnetic configurations. [1] L. Chacon, Computer Physics Communications 163, 143 (2004). [2] L. Chacon, Phys. Plasmas 15, 056103 (2008). [3] S. Cappello, Plasma Phys. Control. Fusion 46, B313 (2004) & references therein.

Modular coils and finite-β operation of a quasi-axially symmetric tokamak

NASA Astrophysics Data System (ADS)

Drevlak, M.

1998-09-01

Quasi-axially symmetric tokamaks (QA tokamaks) are an extension of the conventional tokamak concept. In these devices the magnetic field strength is independent of the generalized toroidal magnetic co-ordinate even though the cross-sectional shape changes. An optimized plasma equilibrium belonging to the class of QA tokamaks has been proposed by Nührenberg. It features the small aspect ratio of a tokamak while allowing part of the rotational transform to be generated by the external field. In this article, two particular aspects of the viability of QA tokamaks are explored, namely the feasibility of modular coils and the possibility of maintaining quasi-axial symmetry in the free-boundary equilibria obtained with the coils found. A set of easily feasible modular coils for the configuration is presented. It was designed using the extended version of the NESCOIL code (Merkel, P., Nucl. Fusion 27 (1987) 867). Using this coil system, free-boundary calculations of the plasma equilibrium were carried out using the NEMEC code (Hirshman, S.P., Van Rij, W.I., Merkel, P., Comput. Phys. Commun. 43 (1986) 143). It is observed that the effects of finite β and net toroidal plasma current can be compensated for with good precision by applying a vertical magnetic field and by separately adjusting the currents of the modular coils. A set of fully three dimensional (3-D) auxiliary coils is proposed to exert control on the rotational transform in the plasma. Deterioration of the quasi-axial symmetry induced by the auxiliary coils can be avoided by adequate adjustment of the currents in the primary coils. Finally, the neoclassical transport properties of the configuration are examined. It is observed that optimization with respect to confinement of the alpha particles can be maintained at operation with finite toroidal current if the aforementioned corrective measures are used. In this case, the neoclassical behaviour is shown to be very similar to that of a conventional tokamak.

Investigating the radial structure of axisymmetric fluctuations in the TCV tokamak with local and global gyrokinetic GENE simulations

NASA Astrophysics Data System (ADS)

Merlo, G.; Brunner, S.; Huang, Z.; Coda, S.; Görler, T.; Villard, L.; Bañón Navarro, A.; Dominski, J.; Fontana, M.; Jenko, F.; Porte, L.; Told, D.

2018-03-01

Axisymmetric (n = 0) density fluctuations measured in the TCV tokamak are observed to possess a frequency f 0 which is either varying (radially dispersive oscillations) or a constant over a large fraction of the plasma minor radius (radially global oscillations) as reported in a companion paper (Z Huang et al, this issue). Given that f 0 scales with the sound speed and given the poloidal structure of density fluctuations, these oscillations were interpreted as Geodesic Acoustic Modes, even though f 0 is in fact smaller than the local linear GAM frequency {f}{GAM}. In this work we employ the Eulerian gyrokinetic code GENE to simulate TCV relevant conditions and investigate the nature and properties of these oscillations, in particular their relation to the safety factor profile. Local and global simulations are carried out and a good qualitative agreement is observed between experiments and simulations. By varying also the plasma temperature and density profiles, we conclude that a variation of the edge safety factor alone is not sufficient to induce a transition from global to radially inhomogeneous oscillations, as was initially suggested by experimental results. This transition appears instead to be the combined result of variations in the different plasma profiles, collisionality and finite machine size effects. Simulations also show that radially global GAM-like oscillations can be observed in all fluxes and fluctuation fields, suggesting that they are the result of a complex nonlinear process involving also finite toroidal mode numbers and not just linear global GAM eigenmodes.

ITG-TEM turbulence simulation with bounce-averaged kinetic electrons in tokamak geometry

NASA Astrophysics Data System (ADS)

Kwon, Jae-Min; Qi, Lei; Yi, S.; Hahm, T. S.

2017-06-01

We develop a novel numerical scheme to simulate electrostatic turbulence with kinetic electron responses in magnetically confined toroidal plasmas. Focusing on ion gyro-radius scale turbulences with slower frequencies than the time scales for electron parallel motions, we employ and adapt the bounce-averaged kinetic equation to model trapped electrons for nonlinear turbulence simulation with Coulomb collisions. Ions are modeled by employing the gyrokinetic equation. The newly developed scheme is implemented on a global δf particle in cell code gKPSP. By performing linear and nonlinear simulations, it is demonstrated that the new scheme can reproduce key physical properties of Ion Temperature Gradient (ITG) and Trapped Electron Mode (TEM) instabilities, and resulting turbulent transport. The overall computational cost of kinetic electrons using this novel scheme is limited to 200%-300% of the cost for simulations with adiabatic electrons. Therefore the new scheme allows us to perform kinetic simulations with trapped electrons very efficiently in magnetized plasmas.

User's manual for the time-dependent INERTIA code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bailey, A.W.; Bennett, R.B.

1985-01-01

The time-dependent INERTIA code is described. This code models the effects of neutral beam momentum input in tokamaks as predicted by the time-dependent formulation of the Stacey-Sigmar formalism. The operation and architecture of the code are described, as are the supplementary plotting and impurity line radiation routines. A short description of the steady-state version of the INERTIA code is also provided.

3D Field Modifications of Core Neutral Fueling In the EMC3-EIRENE Code

NASA Astrophysics Data System (ADS)

Waters, Ian; Frerichs, Heinke; Schmitz, Oliver; Ahn, Joon-Wook; Canal, Gustavo; Evans, Todd; Feng, Yuehe; Kaye, Stanley; Maingi, Rajesh; Soukhanovskii, Vsevolod

2017-10-01

The application of 3-D magnetic field perturbations to the edge plasmas of tokamaks has long been seen as a viable way to control damaging Edge Localized Modes (ELMs). These 3-D fields have also been correlated with a density drop in the core plasmas of tokamaks; known as `pump-out'. While pump-out is typically explained as the result of enhanced outward transport, degraded fueling of the core may also play a role. By altering the temperature and density of the plasma edge, 3-D fields will impact the distribution function of high energy neutral particles produced through ion-neutral energy exchange processes. Starved of the deeply penetrating neutral source, the core density will decrease. Numerical studies carried out with the EMC3-EIRENE code on National Spherical Tokamak eXperiment-Upgrade (NSTX-U) equilibria show that this change to core fueling by high energy neutrals may be a significant contributor to the overall particle balance in the NSTX-U tokamak: deep core (Ψ < 0.5) fueling from neutral ionization sources is decreased by 40-60% with RMPs. This work was funded by the US Department of Energy under Grant DE-SC0012315.

Coupled two-dimensional edge plasma and neutral gas modeling of tokamak scrape-off-layers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maingi, Rajesh

1992-08-01

The objective of this study is to devise a detailed description of the tokamak scrape-off-layer (SOL), which includes the best available models of both the plasma and neutral species and the strong coupling between the two in many SOL regimes. A good estimate of both particle flux and heat flux profiles at the limiter/divertor target plates is desired. Peak heat flux is one of the limiting factors in determining the survival probability of plasma-facing-components at high power levels. Plate particle flux affects the neutral flux to the pump, which determines the particle exhaust rate. A technique which couples a two-dimensionalmore » (2-D) plasma and a 2-D neutral transport code has been developed (coupled code technique), but this procedure requires large amounts of computer time. Relevant physics has been added to an existing two-neutral-species model which takes the SOL plasma/neutral coupling into account in a simple manner (molecular physics model), and this model is compared with the coupled code technique mentioned above. The molecular physics model is benchmarked against experimental data from a divertor tokamak (DIII-D), and a similar model (single-species model) is benchmarked against data from a pump-limiter tokamak (Tore Supra). The models are then used to examine two key issues: free-streaming-limits (ion energy conduction and momentum flux) and the effects of the non-orthogonal geometry of magnetic flux surfaces and target plates on edge plasma parameter profiles.« less

The Physics of Local Helicity Injection Non-Solenoidal Tokamak Startup

NASA Astrophysics Data System (ADS)

Redd, A. J.; Barr, J. L.; Bongard, M. W.; Fonck, R. J.; Hinson, E. T.; Jardin, S.

2013-10-01

Non-solenoidal startup via Local Helicity Injection (LHI) uses compact current injectors to produce toroidal plasma current Ip up to 170 kA in the PEGASUS Toroidal Experiment, driven by 4-8 kA injector current on timescales of 5-20 milliseconds. Increasing the Ip buildup duration enables experimental demonstration of plasma position control on timescales relevant for high-current startup. LHI-driven discharges exhibit bursty MHD activity, apparently line-tied kinking of LHI-driven field lines, with the bursts correlating with rapid equilibrium changes, sharp Ip rises, and sharp drops in the injector impedance. Preliminary NIMROD results suggest that helical LHI-driven current channels remain coherent, with Ip increases due to reconnection between adjacent helical turns forming axisymmetric plasmoids, and corresponding sharp drops in the bias circuit impedance. The DC injector impedance is consistent with a space charge limit at low bias current and a magnetic limit at high bias current. Internal measurements show the current density profile starts strongly hollow and rapidly fills in during Ip buildup. Simulations of LHI discharges using the Tokamak Simulation Code (TSC) will provide insight into the detailed current drive mechanism and guide experiments on PEFASUS and NSTX-U. Work supported by US DOE Grants DE-FG02-96ER54375 and DE-SC0006928.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Podestà, M., E-mail: mpodesta@pppl.gov; Gorelenkova, M.; Fredrickson, E. D.

Integrated simulations of tokamak discharges typically rely on classical physics to model energetic particle (EP) dynamics. However, there are numerous cases in which energetic particles can suffer additional transport that is not classical in nature. Examples include transport by applied 3D magnetic perturbations and, more notably, by plasma instabilities. Focusing on the effects of instabilities, ad-hoc models can empirically reproduce increased transport, but the choice of transport coefficients is usually somehow arbitrary. New approaches based on physics-based reduced models are being developed to address those issues in a simplified way, while retaining a more correct treatment of resonant wave-particle interactions.more » The kick model implemented in the tokamak transport code TRANSP is an example of such reduced models. It includes modifications of the EP distribution by instabilities in real and velocity space, retaining correlations between transport in energy and space typical of resonant EP transport. The relevance of EP phase space modifications by instabilities is first discussed in terms of predicted fast ion distribution. Results are compared with those from a simple, ad-hoc diffusive model. It is then shown that the phase-space resolved model can also provide additional insight into important issues such as internal consistency of the simulations and mode stability through the analysis of the power exchanged between energetic particles and the instabilities.« less

Fully Nonlinear Edge Gyrokinetic Simulations of Kinetic Geodesic-Acoustic Modes and Boundary Flows

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, X Q; Belli, E; Bodi, K

We present edge gyrokinetic neoclassical simulations of tokamak plasmas using the fully nonlinear (full-f) continuum code TEMPEST. A nonlinear Boltzmann model is used for the electrons. The electric field is obtained by solving the 2D gyrokinetic Poisson Equation. We demonstrate the following: (1) High harmonic resonances (n > 2) significantly enhance geodesic-acoustic mode (GAM) damping at high-q (tokamak safety factor), and are necessary to explain both the damping observed in our TEMPEST q-scans and experimental measurements of the scaling of the GAM amplitude with edge q{sub 95} in the absence of obvious evidence that there is a strong q dependencemore » of the turbulent drive and damping of the GAM. (2) The kinetic GAM exists in the edge for steep density and temperature gradients in the form of outgoing waves, its radial scale is set by the ion temperature profile, and ion temperature inhomogeneity is necessary for GAM radial propagation. (3) The development of the neoclassical electric field evolves through different phases of relaxation, including GAMs, their radial propagation, and their long-time collisional decay. (4) Natural consequences of orbits in the pedestal and scrape-off layer region in divertor geometry are substantial non-Maxwellian ion distributions and flow characteristics qualitatively like those observed in experiments.« less

A novel flexible field-aligned coordinate system for tokamak edge plasma simulation

NASA Astrophysics Data System (ADS)

Leddy, J.; Dudson, B.; Romanelli, M.; Shanahan, B.; Walkden, N.

2017-03-01

Tokamak plasmas are confined by a magnetic field that limits the particle and heat transport perpendicular to the field. Parallel to the field the ionised particles can move freely, so to obtain confinement the field lines are "closed" (i.e. form closed surfaces of constant poloidal flux) in the core of a tokamak. Towards, the edge, however, the field lines intersect physical surfaces, leading to interaction between neutral and ionised particles, and the potential melting of the material surface. Simulation of this interaction is important for predicting the performance and lifetime of future tokamak devices such as ITER. Field-aligned coordinates are commonly used in the simulation of tokamak plasmas due to the geometry and magnetic topology of the system. However, these coordinates are limited in the geometry they allow in the poloidal plane due to orthogonality requirements. A novel 3D coordinate system is proposed herein that relaxes this constraint so that any arbitrary, smoothly varying geometry can be matched in the poloidal plane while maintaining a field-aligned coordinate. This system is implemented in BOUT++ and tested for accuracy using the method of manufactured solutions. A MAST edge cross-section is simulated using a fluid plasma model and the results show expected behaviour for density, temperature, and velocity. Finally, simulations of an isolated divertor leg are conducted with and without neutrals to demonstrate the ion-neutral interaction near the divertor plate and the corresponding beneficial decrease in plasma temperature.

The EPQ Code System for Simulating the Thermal Response of Plasma-Facing Components to High-Energy Electron Impact

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ward, Robert Cameron; Steiner, Don

2004-06-15

The generation of runaway electrons during a thermal plasma disruption is a concern for the safe and economical operation of a tokamak power system. Runaway electrons have high energy, 10 to 300 MeV, and may potentially cause extensive damage to plasma-facing components (PFCs) through large temperature increases, melting of metallic components, surface erosion, and possible burnout of coolant tubes. The EPQ code system was developed to simulate the thermal response of PFCs to a runaway electron impact. The EPQ code system consists of several parts: UNIX scripts that control the operation of an electron-photon Monte Carlo code to calculate themore » interaction of the runaway electrons with the plasma-facing materials; a finite difference code to calculate the thermal response, melting, and surface erosion of the materials; a code to process, scale, transform, and convert the electron Monte Carlo data to volumetric heating rates for use in the thermal code; and several minor and auxiliary codes for the manipulation and postprocessing of the data. The electron-photon Monte Carlo code used was Electron-Gamma-Shower (EGS), developed and maintained by the National Research Center of Canada. The Quick-Therm-Two-Dimensional-Nonlinear (QTTN) thermal code solves the two-dimensional cylindrical modified heat conduction equation using the Quickest third-order accurate and stable explicit finite difference method and is capable of tracking melting or surface erosion. The EPQ code system is validated using a series of analytical solutions and simulations of experiments. The verification of the QTTN thermal code with analytical solutions shows that the code with the Quickest method is better than 99.9% accurate. The benchmarking of the EPQ code system and QTTN versus experiments showed that QTTN's erosion tracking method is accurate within 30% and that EPQ is able to predict the occurrence of melting within the proper time constraints. QTTN and EPQ are verified and validated as able to calculate the temperature distribution, phase change, and surface erosion successfully.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strauss, H.R.

This paper describes the code FEMHD, an adaptive finite element MHD code, which is applied in a number of different manners to model MHD behavior and edge plasma phenomena on a diverted tokamak. The code uses an unstructured triangular mesh in 2D and wedge shaped mesh elements in 3D. The code has been adapted to look at neutral and charged particle dynamics in the plasma scrape off region, and into a full MHD-particle code.

Active stabilization of error field penetration via control field and bifurcation of its stable frequency range

NASA Astrophysics Data System (ADS)

Inoue, S.; Shiraishi, J.; Takechi, M.; Matsunaga, G.; Isayama, A.; Hayashi, N.; Ide, S.

2017-11-01

An active stabilization effect of a rotating control field against an error field penetration is numerically studied. We have developed a resistive magnetohydrodynamic code ‘AEOLUS-IT’, which can simulate plasma responses to rotating/static external magnetic field. Adopting non-uniform flux coordinates system, the AEOLUS-IT simulation can employ high magnetic Reynolds number condition relevant to present tokamaks. By AEOLUS-IT, we successfully clarified the stabilization mechanism of the control field against the error field penetration. Physical processes of a plasma rotation drive via the control field are demonstrated by the nonlinear simulation, which reveals that the rotation amplitude at a resonant surface is not a monotonic function of the control field frequency, but has an extremum. Consequently, two ‘bifurcated’ frequency ranges of the control field are found for the stabilization of the error field penetration.

Performance of JT-60SA divertor Thomson scattering diagnostics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kajita, Shin, E-mail: kajita.shin@nagoya-u.jp; Hatae, Takaki; Tojo, Hiroshi

2015-08-15

For the satellite tokamak JT-60 Super Advanced (JT-60SA), a divertor Thomson scattering measurement system is planning to be installed. In this study, we improved the design of the collection optics based on the previous one, in which it was found that the solid angle of the collection optics became very small, mainly because of poor accessibility to the measurement region. By improvement, the solid angle was increased by up to approximately five times. To accurately assess the measurement performance, background noise was assessed using the plasma parameters in two typical discharges in JT-60SA calculated from the SONIC code. Moreover, themore » influence of the reflection of bremsstrahlung radiation by the wall is simulated by using a ray tracing simulation. The errors in the temperature and the density are assessed based on the simulation results for three typical field of views.« less

Performance of JT-60SA divertor Thomson scattering diagnostics.

PubMed

Kajita, Shin; Hatae, Takaki; Tojo, Hiroshi; Enokuchi, Akito; Hamano, Takashi; Shimizu, Katsuhiro; Kawashima, Hisato

2015-08-01

For the satellite tokamak JT-60 Super Advanced (JT-60SA), a divertor Thomson scattering measurement system is planning to be installed. In this study, we improved the design of the collection optics based on the previous one, in which it was found that the solid angle of the collection optics became very small, mainly because of poor accessibility to the measurement region. By improvement, the solid angle was increased by up to approximately five times. To accurately assess the measurement performance, background noise was assessed using the plasma parameters in two typical discharges in JT-60SA calculated from the SONIC code. Moreover, the influence of the reflection of bremsstrahlung radiation by the wall is simulated by using a ray tracing simulation. The errors in the temperature and the density are assessed based on the simulation results for three typical field of views.

Implementation of non-axisymmetric mesh system in the gyrokinetic PIC code (XGC) for Stellarators

NASA Astrophysics Data System (ADS)

Moritaka, Toseo; Hager, Robert; Cole, Micheal; Chang, Choong-Seock; Lazerson, Samuel; Ku, Seung-Hoe; Ishiguro, Seiji

2017-10-01

Gyrokinetic simulation is a powerful tool to investigate turbulent and neoclassical transports based on the first-principles of plasma kinetics. The gyrokinetic PIC code XGC has been developed for integrated simulations that cover the entire region of Tokamaks. Complicated field line and boundary structures should be taken into account to demonstrate edge plasma dynamics under the influence of X-point and vessel components. XGC employs gyrokinetic Poisson solver on unstructured triangle mesh to deal with this difficulty. We introduce numerical schemes newly developed for XGC simulation in non-axisymmetric Stellarator geometry. Triangle meshes in each poloidal plane are defined by PEST poloidal angle in the VMEC equilibrium so that they have the same regular structure in the straight field line coordinate. Electric charge of marker particle is distributed to the triangles specified by the field-following projection to the neighbor poloidal planes. 3D spline interpolation in a cylindrical mesh is also used to obtain equilibrium magnetic field at the particle position. These schemes capture the anisotropic plasma dynamics and resulting potential structure with high accuracy. The triangle meshes can smoothly connect to unstructured meshes in the edge region. We will present the validation test in the core region of Large Helical Device and discuss about future challenges toward edge simulations.

Investigation of the transport shortfall in Alcator C-Mod L-mode plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howard, N. T.; White, A. E.; Greenwald, M.

2013-03-15

A so-called 'transport shortfall,' where ion and electron heat fluxes and turbulence are underpredicted by gyrokinetic codes, has been robustly identified in DIII-D L-mode plasmas for {rho}>0.55[T. L. Rhodes et al., Nucl. Fusion 51(6), 063022 (2011); and C. Holland et al., Phys. Plasmas 16(5), 052301 (2009)]. To probe the existence of a transport shortfall across different tokamaks, a dedicated scan of auxiliary heated L-mode discharges in Alcator C-Mod are studied in detail with nonlinear gyrokinetic simulations for the first time. Two discharges, only differing by the amount of auxiliary heating are investigated using both linear and nonlinear simulation of themore » GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)]. Nonlinear gyrokinetic simulation of the low and high input power discharges reveals a discrepancy between simulation and experiment in only the electron heat flux channel of the low input power discharge. However, both discharges demonstrate excellent agreement in the ion heat flux channel, and the high input power discharge demonstrates simultaneous agreement with experiment in both the electron and ion heat flux channels. A summary of linear and nonlinear gyrokinetic results and a discussion of possible explanations for the agreement/disagreement in each heat flux channel is presented.« less

Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

NASA Astrophysics Data System (ADS)

Deng, Guozhong; Liu, Xiaoju; Wang, Liang; Liu, Shaocheng; Xu, Jichan; Feng, Wei; Liu, Jianbin; Liu, Huan; Gao, Xiang

2017-04-01

The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak (EAST) L-mode and ELM-free H-mode plasmas. The divertor power footprint widths, which consist of the scrape-off layer (SOL) width λ q and heat spreading S, are important physical parameters for edge plasmas. In this work, a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current I p. Strong inverse scaling of the SOL width with I p has been achieved for both L-mode and H-mode plasmas in the forms of {λ }q,{{L}\\text-\\text{mode}}=4.98× {I}{{p}}-0.68 and {λ }q,{{H}\\text-\\text{mode}}=1.86× {I}{{p}}-1.08. Similar trends have also been demonstrated in the study of heat spreading with {S}{{L}\\text-\\text{mode}}=1.95× {I}{{p}}-0.542 and {S}{{H}\\text-\\text{mode}}=0.756× {I}{{p}}-0.872. In addition, studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current. The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor (CFETR).

Combined magnetic and kinetic control of advanced tokamak steady state scenarios based on semi-empirical modelling

NASA Astrophysics Data System (ADS)

Moreau, D.; Artaud, J. F.; Ferron, J. R.; Holcomb, C. T.; Humphreys, D. A.; Liu, F.; Luce, T. C.; Park, J. M.; Prater, R.; Turco, F.; Walker, M. L.

2015-06-01

This paper shows that semi-empirical data-driven models based on a two-time-scale approximation for the magnetic and kinetic control of advanced tokamak (AT) scenarios can be advantageously identified from simulated rather than real data, and used for control design. The method is applied to the combined control of the safety factor profile, q(x), and normalized pressure parameter, βN, using DIII-D parameters and actuators (on-axis co-current neutral beam injection (NBI) power, off-axis co-current NBI power, electron cyclotron current drive power, and ohmic coil). The approximate plasma response model was identified from simulated open-loop data obtained using a rapidly converging plasma transport code, METIS, which includes an MHD equilibrium and current diffusion solver, and combines plasma transport nonlinearity with 0D scaling laws and 1.5D ordinary differential equations. The paper discusses the results of closed-loop METIS simulations, using the near-optimal ARTAEMIS control algorithm (Moreau D et al 2013 Nucl. Fusion 53 063020) for steady state AT operation. With feedforward plus feedback control, the steady state target q-profile and βN are satisfactorily tracked with a time scale of about 10 s, despite large disturbances applied to the feedforward powers and plasma parameters. The robustness of the control algorithm with respect to disturbances of the H&CD actuators and of plasma parameters such as the H-factor, plasma density and effective charge, is also shown.

Investigating the radial structure of axisymmetric fluctuations in the TCV tokamak with local and global gyrokinetic GENE simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Merlo, Gabriele; Brunner, Stephan; Huang, Zhouji

Axisymmetric (n=0) density fluctuations measured in the TCV tokamak are observed to possess a frequency f₀ which is either varying (radially dispersive oscillations) or a constant over a large fraction of the plasma minor radius (radially global oscillations) as reported in a companion paper [Z. Huang et al., this issue]. Given that f₀ scales with the sound speed and given the poloidal structure of density fluctuations, these oscillations were interpreted as Geodesic Acoustic Modes, even though f₀ is in fact smaller than the local linear GAM frequency f_GAM . In this work we employ the Eulerian gyrokinetic code GENE tomore » simulate TCV relevant conditions and investigate the nature properties of these oscillations, in particular their relation to the safety factor profile. Local and global simulations are carried out and a good qualitative agreement is observed between experiments and simulations. By varying also the plasma temperature and density profiles, we conclude that a variation of the edge safety factor alone is not sufficient to induce a transition from global to radially inhomogeneous oscillations, as was initially suggested by experimental results. This transition appears instead to be the combined result of variations in the different plasma profiles, collisionality and finite machine size effects. In conclusion, simulations also show that radially global GAM-like oscillations can be observed in all fluxes and fluctuation fields, suggesting that they are the result of a complex nonlinear process involving also finite toroidal mode numbers and not just linear global GAM eigenmodes.« less

Synergistic cross-scale coupling of turbulence in a tokamak plasma

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howard, N. T., E-mail: nthoward@psfc.mit.edu; Holland, C.; White, A. E.

2014-11-15

For the first time, nonlinear gyrokinetic simulations spanning both the ion and electron spatio-temporal scales have been performed with realistic electron mass ratio ((m{sub D}∕m{sub e}){sup 1∕2 }= 60.0), realistic geometry, and all experimental inputs, demonstrating the coexistence and synergy of ion (k{sub θ}ρ{sub s}∼O(1.0)) and electron-scale (k{sub θ}ρ{sub e}∼O(1.0)) turbulence in the core of a tokamak plasma. All multi-scale simulations utilized the GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] to study the coupling of ion and electron-scale turbulence in the core (r/a = 0.6) of an Alcator C-Mod L-mode discharge shown previously to exhibit an under-predictionmore » of the electron heat flux when using simulations only including ion-scale turbulence. Electron-scale turbulence is found to play a dominant role in setting the electron heat flux level and radially elongated (k{sub r} ≪ k{sub θ}) “streamers” are found to coexist with ion-scale eddies in experimental plasma conditions. Inclusion of electron-scale turbulence in these simulations is found to increase both ion and electron heat flux levels by enhancing the transport at the ion-scale while also driving electron heat flux at sub-ρ{sub i} scales. The combined increases in the low and high-k driven electron heat flux may explain previously observed discrepancies between simulated and experimental electron heat fluxes and indicates a complex interaction of short and long wavelength turbulence.« less

Investigating the radial structure of axisymmetric fluctuations in the TCV tokamak with local and global gyrokinetic GENE simulations

DOE PAGES

Merlo, Gabriele; Brunner, Stephan; Huang, Zhouji; ...

2017-12-19

Axisymmetric (n=0) density fluctuations measured in the TCV tokamak are observed to possess a frequency f₀ which is either varying (radially dispersive oscillations) or a constant over a large fraction of the plasma minor radius (radially global oscillations) as reported in a companion paper [Z. Huang et al., this issue]. Given that f₀ scales with the sound speed and given the poloidal structure of density fluctuations, these oscillations were interpreted as Geodesic Acoustic Modes, even though f₀ is in fact smaller than the local linear GAM frequency f_GAM . In this work we employ the Eulerian gyrokinetic code GENE tomore » simulate TCV relevant conditions and investigate the nature properties of these oscillations, in particular their relation to the safety factor profile. Local and global simulations are carried out and a good qualitative agreement is observed between experiments and simulations. By varying also the plasma temperature and density profiles, we conclude that a variation of the edge safety factor alone is not sufficient to induce a transition from global to radially inhomogeneous oscillations, as was initially suggested by experimental results. This transition appears instead to be the combined result of variations in the different plasma profiles, collisionality and finite machine size effects. In conclusion, simulations also show that radially global GAM-like oscillations can be observed in all fluxes and fluctuation fields, suggesting that they are the result of a complex nonlinear process involving also finite toroidal mode numbers and not just linear global GAM eigenmodes.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ku, S.; Chang, C. S.; Hager, R.

Here, a fast edge turbulence suppression event has been simulated in the electrostatic version of the gyrokinetic particle-in-cell code XGC1 in a realistic diverted tokamak edge geometry under neutral particle recycling. The results show that the sequence of turbulent Reynolds stress followed by neoclassical ion orbit-loss driven together conspire to form the sustaining radial electric field shear and to quench turbulent transport just inside the last closed magnetic flux surface. As a result, the main suppression action is located in a thin radial layer around ψ N≃0.96–0.98, where ψ N is the normalized poloidal flux, with the time scale ~0.1more » ms.« less

EBW H&CD Potential for Spherical Tokamaks

NASA Astrophysics Data System (ADS)

Urban, J.; Decker, J.; Peysson, Y.; Preinhaelter, J.; Shevchenko, V.; Taylor, G.; Vahala, L.; Vahala, G.

2011-12-01

Spherical tokamaks (STs), which feature relatively high neutron flux and good economy, operate generally in high-ß regimes, in which the usual EC O- and X- modes are cut-off. In this case, electron Bernstein waves (EBWs) seem to be the only option that can provide features similar to the EC waves—controllable localized heating and current drive (H&) that can be utilized for core plasma heating as well as for accurate plasma stabilization. We first derive an analytical expression for Gaussian beam OXB conversion efficiency. Then, an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX) is performed. Coupled ray-tracing (AMR) and Fokker-Planck (LUKE) codes are employed to simulate EBWs of varying frequencies and launch conditions. Our results indicate that an efficient and universal EBW H&CD system is indeed viable. In particular, power can be deposited and current reasonably efficiently driven across the whole plasma radius. Such a system could be controlled by a suitably chosen launching antenna vertical position and would also be sufficiently robust.

Simulation of the alpha particle heating and the helium ash source in an International Thermonuclear Experimental Reactor-like tokamak with an internal transport barrier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ye, Lei, E-mail: lye@ipp.ac.cn; Guo, Wenfeng; Xiao, Xiaotao

2014-12-15

A guiding center orbit following code, which incorporates a set of non-singular coordinates for orbit integration, was developed and applied to investigate the alpha particle heating in an ITER-like tokamak with an internal transport barrier. It is found that a relatively large q (safety factor) value can significantly broaden the alpha heating profile in comparison with the local heating approximation; this broadening is due to the finite orbit width effects; when the orbit width is much smaller than the scale length of the alpha particle source profile, the heating profile agrees with the source profile, otherwise, the heating profile canmore » be significantly broadened. It is also found that the stagnation particles move to the magnetic axis during the slowing-down process, thus the effect of stagnation orbits is not beneficial to the helium ash removal. The source profile of helium ash is broadened in comparison with the alpha source profile, which is similar to the heating profile.« less

Dynamics of tokamak plasma surface current in 3D ideal MHD model

NASA Astrophysics Data System (ADS)

Galkin, Sergei A.; Svidzinski, V. A.; Zakharov, L. E.

2013-10-01

Interest in the surface current which can arise on perturbed sharp plasma vacuum interface in tokamaks was recently generated by a few papers (see and references therein). In dangerous disruption events with plasma-touching-wall scenarios, the surface current can be shared with the wall leading to the strong, damaging forces acting on the wall A relatively simple analytic definition of δ-function surface current proportional to a jump of tangential component of magnetic field nevertheless leads to a complex computational problem on the moving plasma-vacuum interface, requiring the incorporation of non-linear 3D plasma dynamics even in one-fluid ideal MHD. The Disruption Simulation Code (DSC), which had recently been developed in a fully 3D toroidal geometry with adaptation to the moving plasma boundary, is an appropriate tool for accurate self-consistent δfunction surface current calculation. Progress on the DSC-3D development will be presented. Self-consistent surface current calculation under non-linear dynamics of low m kink mode and VDE will be discussed. Work is supported by the US DOE SBIR grant #DE-SC0004487.

Radioactivity evaluation for the KSTAR tokamak.

PubMed

Kim, Hyunduk; Lee, Hee-Seock; Hong, Sukmo; Kim, Minho; Chung, Chinwha; Kim, Changsuk

2005-01-01

The deuterium-deuterium (D-D) reaction in the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak generates neutrons with a peak yield of 2.5 x 10(16) s(-1) through a pulse operation of 300 s. Since the structure material of the tokamak is irradiated with neutrons, this environment will restrict work around and inside the tokamak from a radiation protection physics point of view after shutdown. Identification of neutron-produced radionuclides and evaluation of absorbed dose in the structure material are needed to develop a guiding principle for radiation protection. The activation level was evaluated by MCNP4C2 and an inventory code, FISPACT. The absorbed dose in the working area decreased by 4.26 x 10(-4) mrem h(-1) in the inner vessel 1.5 d after shutdown. Furthermore, tritium strongly contributes to the contamination in the graphite tile. The amount of tritium produced by neutrons was 3.03 x 10(6) Bq kg(-1) in the carbon graphite of a plasma-facing wall.

Comparing simulation of plasma turbulence with experiment

NASA Astrophysics Data System (ADS)

Ross, David W.; Bravenec, Ronald V.; Dorland, William; Beer, Michael A.; Hammett, G. W.; McKee, George R.; Fonck, Raymond J.; Murakami, Masanori; Burrell, Keith H.; Jackson, Gary L.; Staebler, Gary M.

2002-01-01

The direct quantitative correspondence between theoretical predictions and the measured plasma fluctuations and transport is tested by performing nonlinear gyro-Landau-fluid simulations with the GRYFFIN (or ITG) code [W. Dorland and G. W. Hammett, Phys. Fluids B 5, 812 (1993); M. A. Beer and G. W. Hammett, Phys. Plasmas 3, 4046 (1996)]. In an L-mode reference discharge in the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)], which has relatively large fluctuations and transport, the turbulence is dominated by ion temperature gradient (ITG) modes. Trapped electron modes and impurity drift waves also play a role. Density fluctuations are measured by beam emission spectroscopy [R. J. Fonck, P. A. Duperrex, and S. F. Paul, Rev. Sci. Instrum. 61, 3487 (1990)]. Experimental fluxes and corresponding diffusivities are analyzed by the TRANSP code [R. J. Hawryluk, in Physics of Plasmas Close to Thermonuclear Conditions, edited by B. Coppi, G. G. Leotta, D. Pfirsch, R. Pozzoli, and E. Sindoni (Pergamon, Oxford, 1980), Vol. 1, p. 19]. The shape of the simulated wave number spectrum is close to the measured one. The simulated ion thermal transport, corrected for E×B low shear, exceeds the experimental value by a factor of 1.5 to 2.0. The simulation overestimates the density fluctuation level by an even larger factor. On the other hand, the simulation underestimates the electron thermal transport, which may be accounted for by modes that are not accessible to the simulation or to the BES measurement.

Feedback control design for non-inductively sustained scenarios in NSTX-U using TRANSP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyer, M. D.; Andre, R. G.; Gates, D. A.

This paper examines a method for real-time control of non-inductively sustained scenarios in NSTX-U by using TRANSP, a time-dependent integrated modeling code for prediction and interpretive analysis of tokamak experimental data, as a simulator. The actuators considered for control in this work are the six neutral beam sources and the plasma boundary shape. To understand the response of the plasma current, stored energy, and central safety factor to these actuators and to enable systematic design of control algorithms, simulations were run in which the actuators were modulated and a linearized dynamic response model was generated. A multi-variable model-based control schememore » that accounts for the coupling and slow dynamics of the system while mitigating the effect of actuator limitations was designed and simulated. Simulations show that modest changes in the outer gap and heating power can improve the response time of the system, reject perturbations, and track target values of the controlled values.« less

Feedback control design for non-inductively sustained scenarios in NSTX-U using TRANSP

DOE PAGES

Boyer, M. D.; Andre, R. G.; Gates, D. A.; ...

2017-04-24

This paper examines a method for real-time control of non-inductively sustained scenarios in NSTX-U by using TRANSP, a time-dependent integrated modeling code for prediction and interpretive analysis of tokamak experimental data, as a simulator. The actuators considered for control in this work are the six neutral beam sources and the plasma boundary shape. To understand the response of the plasma current, stored energy, and central safety factor to these actuators and to enable systematic design of control algorithms, simulations were run in which the actuators were modulated and a linearized dynamic response model was generated. A multi-variable model-based control schememore » that accounts for the coupling and slow dynamics of the system while mitigating the effect of actuator limitations was designed and simulated. Simulations show that modest changes in the outer gap and heating power can improve the response time of the system, reject perturbations, and track target values of the controlled values.« less

Feedback control design for non-inductively sustained scenarios in NSTX-U using TRANSP

NASA Astrophysics Data System (ADS)

Boyer, M. D.; Andre, R. G.; Gates, D. A.; Gerhardt, S. P.; Menard, J. E.; Poli, F. M.

2017-06-01

This paper examines a method for real-time control of non-inductively sustained scenarios in NSTX-U by using TRANSP, a time-dependent integrated modeling code for prediction and interpretive analysis of tokamak experimental data, as a simulator. The actuators considered for control in this work are the six neutral beam sources and the plasma boundary shape. To understand the response of the plasma current, stored energy, and central safety factor to these actuators and to enable systematic design of control algorithms, simulations were run in which the actuators were modulated and a linearized dynamic response model was generated. A multi-variable model-based control scheme that accounts for the coupling and slow dynamics of the system while mitigating the effect of actuator limitations was designed and simulated. Simulations show that modest changes in the outer gap and heating power can improve the response time of the system, reject perturbations, and track target values of the controlled values.

Progress in Development of the ITER Plasma Control System Simulation Platform

NASA Astrophysics Data System (ADS)

Walker, Michael; Humphreys, David; Sammuli, Brian; Ambrosino, Giuseppe; de Tommasi, Gianmaria; Mattei, Massimiliano; Raupp, Gerhard; Treutterer, Wolfgang; Winter, Axel

2017-10-01

We report on progress made and expected uses of the Plasma Control System Simulation Platform (PCSSP), the primary test environment for development of the ITER Plasma Control System (PCS). PCSSP will be used for verification and validation of the ITER PCS Final Design for First Plasma, to be completed in 2020. We discuss the objectives of PCSSP, its overall structure, selected features, application to existing devices, and expected evolution over the lifetime of the ITER PCS. We describe an archiving solution for simulation results, methods for incorporating physics models of the plasma and physical plant (tokamak, actuator, and diagnostic systems) into PCSSP, and defining characteristics of models suitable for a plasma control development environment such as PCSSP. Applications of PCSSP simulation models including resistive plasma equilibrium evolution are demonstrated. PCSSP development supported by ITER Organization under ITER/CTS/6000000037. Resistive evolution code developed under General Atomics' Internal funding. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

Comparing TCV experimental VDE responses with DINA code simulations

NASA Astrophysics Data System (ADS)

Favez, J.-Y.; Khayrutdinov, R. R.; Lister, J. B.; Lukash, V. E.

2002-02-01

The DINA free-boundary equilibrium simulation code has been implemented for TCV, including the full TCV feedback and diagnostic systems. First results showed good agreement with control coil perturbations and correctly reproduced certain non-linear features in the experimental measurements. The latest DINA code simulations, presented in this paper, exploit discharges with different cross-sectional shapes and different vertical instability growth rates which were subjected to controlled vertical displacement events (VDEs), extending previous work with the DINA code on the DIII-D tokamak. The height of the TCV vessel allows observation of the non-linear evolution of the VDE growth rate as regions of different vertical field decay index are crossed. The vertical movement of the plasma is found to be well modelled. For most experiments, DINA reproduces the S-shape of the vertical displacement in TCV with excellent precision. This behaviour cannot be modelled using linear time-independent models because of the predominant exponential shape due to the unstable pole of any linear time-independent model. The other most common equilibrium parameters like the plasma current Ip, the elongation κ, the triangularity δ, the safety factor q, the ratio between the averaged plasma kinetic pressure and the pressure of the poloidal magnetic field at the edge of the plasma βp, and the internal self inductance li also show acceptable agreement. The evolution of the growth rate γ is estimated and compared with the evolution of the closed-loop growth rate calculated with the RZIP linear model, confirming the origin of the observed behaviour.

Final Technical Report for the Center for Momentum Transport and Flow Organization (CMTFO)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forest, Cary B.; Tynan, George R.

The Center for Momentum Transport and Flow Organization (CMTFO) is a DOE Plasma Science Center formed in late 2009 to focus on the general principles underlying momentum transport in magnetic fusion and astrophysical systems. It is composed of funded researchers from UCSD, UW Madison, U. Colorado, PPPL. As of 2011, UCSD supported postdocs are collaborating at MIT/Columbia and UC Santa Cruz and beginning in 2012, will also be based at PPPL. In the initial startup period, the Center supported the construction of two basic experiments at PPPL and UW Madison to focus on accretion disk hydrodynamic instabilities and solar physicsmore » issues. We now have computational efforts underway focused on understanding recent experimental tests of dynamos, solar tacholine physics, intrinsic rotation in tokamak plasmas and L-H transition physics in tokamak devices. In addition, we have the basic experiments discussed above complemented by work on a basic linear plasma device at UCSD and a collaboration at the LAPD located at UCLA. We are also performing experiments on intrinsic rotation and L-H transition physics in the DIII-D, NSTX, C-Mod, HBT EP, HL-2A, and EAST tokamaks in the US and China, and expect to begin collaborations on K-STAR in the coming year. Center funds provide support to over 10 postdocs and graduate students each year, who work with 8 senior faculty and researchers at their respective institutions. The Center has sponsored a mini-conference at the APS DPP 2010 meeting, and co-sponsored the recent Festival de Theorie (2011) with the CEA in Cadarache, and will co-sponsor a Winter School in January 2012 in collaboration with the CMSO-UW Madison. Center researchers have published over 50 papers in the peer reviewed literature, and given over 10 talks at major international meetings. In addition, the Center co-PI, Professor Patrick Diamond, shared the 2011 Alfven Prize at the EPS meeting. Key scientific results from this startup period include initial simulations of the effects of boundary conditions on turbulent dynamo experiments; simulations of intrinsic rotation showing the strong link between toroidal rotation and temperature gradients and elucidation of the turbulence symmetry breaking mechanisms that lead to this macroscopic behavior; first experiments in a large tokamak testing the roll of turbulent momentum transport in driving intrinsic rotation; experiments in tokamaks showing strong evidence that zonal flows, together with the more widely recognized mean sheared ExB flow, act to trigger the L-H transition in tokamak devices and the first experimental measurement of collisional viscosity in an unmagnetized plasma. In the coming three year period, we will continue these efforts by a combination of basic hydrodynamic, liquid metal and plasma experiments combined with experiments on numerous tokamak devices around the world. In addition, we will use MHD, gyrofluid and gyrokinetic codes combined with theory to address the problems of interest to the Center.« less

Center for Momentum Transport and Flow Organization (CMTFO). Final technical report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tynan, George R.; Diamond, P. H.; Ji, H.

The Center for Momentum Transport and Flow Organization (CMTFO) is a DOE Plasma Science Center formed in late 2009 to focus on the general principles underlying momentum transport in magnetic fusion and astrophysical systems. It is composed of funded researchers from UCSD, UW Madison, U. Colorado, PPPL. As of 2011, UCSD supported postdocs are collaborating at MIT/Columbia and UC Santa Cruz and beginning in 2012, will also be based at PPPL. In the initial startup period, the Center supported the construction of two basic experiments at PPPL and UW Madison to focus on accretion disk hydrodynamic instabilities and solar physicsmore » issues. We now have computational efforts underway focused on understanding recent experimental tests of dynamos, solar tachocline physics, intrinsic rotation in tokamak plasmas and L-H transition physics in tokamak devices. In addition, we have the basic experiments discussed above complemented by work on a basic linear plasma device at UCSD and a collaboration at the LAPD located at UCLA. We are also performing experiments on intrinsic rotation and L-H transition physics in the DIII-D, NSTX, C-Mod, HBT EP, HL-2A, and EAST tokamaks in the US and China, and expect to begin collaborations on K-STAR in the coming year. Center funds provide support to over 10 postdocs and graduate students each year, who work with 8 senior faculty and researchers at their respective institutions. The Center has sponsored a mini-conference at the APS DPP 2010 meeting, and co-sponsored the recent Festival de Theorie (2011) with the CEA in Cadarache, and will co-sponsor a Winter School in January 2012 in collaboration with the CMSO-UW Madison. Center researchers have published over 50 papers in the peer reviewed literature, and given over 10 talks at major international meetings. In addition, the Center co-PI, Professor Patrick Diamond, shared the 2011 Alfven Prize at the EPS meeting. Key scientific results from this startup period include initial simulations of the effects of boundary conditions on turbulent dynamo experiments; simulations of intrinsic rotation showing the strong link between toroidal rotation and temperature gradients and elucidation of the turbulence symmetry breaking mechanisms that lead to this macroscopic behavior; first experiments in a large tokamak testing the roll of turbulent momentum transport in driving intrinsic rotation; experiments in tokamaks showing strong evidence that zonal flows, together with the more widely recognized mean sheared ExB flow, act to trigger the L-H transition in tokamak devices and the first experimental measurement of collisional viscosity in an unmagnetized plasma. In the coming three year period, we will continue these efforts by a combination of basic hydrodynamic, liquid metal and plasma experiments combined with experiments on numerous tokamak devices around the world. In addition, we will use MHD, gyrofluid and gyrokinetic codes combined with theory to address the problems of interest to the Center.« less

Development of Tokamak Transport Solvers for Stiff Confinement Systems

NASA Astrophysics Data System (ADS)

St. John, H. E.; Lao, L. L.; Murakami, M.; Park, J. M.

2006-10-01

Leading transport models such as GLF23 [1] and MM95 [2] describe turbulent plasma energy, momentum and particle flows. In order to accommodate existing transport codes and associated solution methods effective diffusivities have to be derived from these turbulent flow models. This can cause significant problems in predicting unique solutions. We have developed a parallel transport code solver, GCNMP, that can accommodate both flow based and diffusivity based confinement models by solving the discretized nonlinear equations using modern Newton, trust region, steepest descent and homotopy methods. We present our latest development efforts, including multiple dynamic grids, application of two-level parallel schemes, and operator splitting techniques that allow us to combine flow based and diffusivity based models in tokamk simulations. 6pt [1] R.E. Waltz, et al., Phys. Plasmas 4, 7 (1997). [2] G. Bateman, et al., Phys. Plasmas 5, 1793 (1998).

Fast particles in a steady-state compact FNS and compact ST reactor

NASA Astrophysics Data System (ADS)

Gryaznevich, M. P.; Nicolai, A.; Buxton, P.

2014-10-01

This paper presents results of studies of fast particles (ions and alpha particles) in a steady-state compact fusion neutron source (CFNS) and a compact spherical tokamak (ST) reactor with Monte-Carlo and Fokker-Planck codes. Full-orbit simulations of fast particle physics indicate that a compact high field ST can be optimized for energy production by a reduction of the necessary (for the alpha containment) plasma current compared with predictions made using simple analytic expressions, or using guiding centre approximation in a numerical code. Alpha particle losses may result in significant heating and erosion of the first wall, so such losses for an ST pilot plant have been calculated and total and peak wall loads dependence on the plasma current has been studied. The problem of dilution has been investigated and results for compact and big size devices are compared.

Limitations of bootstrap current models

DOE PAGES

Belli, Emily A.; Candy, Jefferey M.; Meneghini, Orso; ...

2014-03-27

We assess the accuracy and limitations of two analytic models of the tokamak bootstrap current: (1) the well-known Sauter model and (2) a recent modification of the Sauter model by Koh et al. For this study, we use simulations from the first-principles kinetic code NEO as the baseline to which the models are compared. Tests are performed using both theoretical parameter scans as well as core- to-edge scans of real DIII-D and NSTX plasma profiles. The effects of extreme aspect ratio, large impurity fraction, energetic particles, and high collisionality are studied. In particular, the error in neglecting cross-species collisional couplingmore » – an approximation inherent to both analytic models – is quantified. Moreover, the implications of the corrections from kinetic NEO simulations on MHD equilibrium reconstructions is studied via integrated modeling with kinetic EFIT.« less

Simulation of turbulence in the divertor region of tokamak edge plasma

NASA Astrophysics Data System (ADS)

Umansky, M. V.; Rognlien, T. D.; Xu, X. Q.

2005-03-01

Results are presented for turbulence simulations with the fluid edge turbulence code BOUT [X.Q. Xu, R.H. Cohen, Contr. Plas. Phys. 36 (1998) 158]. The present study is focussed on turbulence in the divertor leg region and on the role of the X-point in the structure of turbulence. Results of the present calculations indicate that the ballooning effects are important for the divertor fluctuations. The X-point shear leads to weak correlation of turbulence across the X-point regions, in particular for large toroidal wavenumber. For the saturated amplitudes of the divertor region turbulence it is found that amplitudes of density fluctuations are roughly proportional to the local density of the background plasma. The amplitudes of electron temperature and electric potential fluctuations are roughly proportional to the local electron temperature of the background plasma.

Simulation of Plasma Transport in a Toroidal Annulus with TEMPEST

NASA Astrophysics Data System (ADS)

Xiong, Z.

2005-10-01

TEMPEST is an edge gyro-kinetic continuum code currently under development at LLNL to study boundary plasma transport over a region extending from inside the H-mode pedestal across the separatrix to the divertor plates. Here we report simulation results from the 4D (θ, ψ, E, μ) TEMPEST, for benchmark purpose, in an annulus region immediately inside the separatrix of a large aspect ratio, circular cross-section tokamak. Besides the normal poloidal trapping regions, there are radial inaccessible regions at a fixed poloid angle, energy and magnetic moment due to the radial variation of the B field. To handle such cases, a fifth-order WENO differencing scheme is used in the radial direction. The particle and heat transport coefficients are obtained for different collisional regimes and compared with the neo-classical transport theory.

Full wave simulations of fast wave efficiency and power losses in the scrape-off layer of tokamak plasmas in mid/high harmonic and minority heating regimes

DOE PAGES

Bertelli, N.; Jaeger, E. F.; Hosea, J. C.; ...

2015-12-17

Here, several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves (HHFW), have found strong interaction between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 2D and 3D AORSAmore » results for the National Spherical Torus eXperiment (NSTX) have shown a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is removed from in front of the antenna by increasing the edge density. Here, full wave simulations have been extended for 'conventional' tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results in HHFW regime show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for C-Mod and EAST, which operate in the minority heating regime.« less

Phase space effects on fast ion distribution function modeling in tokamaks

DOE PAGES

Podesta, M.; Gorelenkova, M.; Fredrickson, E. D.; ...

2016-04-14

Here, integrated simulations of tokamak discharges typically rely on classical physics to model energetic particle (EP) dynamics. However, there are numerous cases in which energetic particles can suffer additional transport that is not classical in nature. Examples include transport by applied 3D magnetic perturbations and, more notably, by plasma instabilities. Focusing on the effects of instabilities,ad-hocmodels can empirically reproduce increased transport, but the choice of transport coefficients is usually somehow arbitrary. New approaches based on physics-based reduced models are being developed to address those issues in a simplified way, while retaining a more correct treatment of resonant wave-particle interactions. Themore » kick model implemented in the tokamaktransport code TRANSP is an example of such reduced models. It includes modifications of the EP distribution by instabilities in real and velocity space, retaining correlations between transport in energy and space typical of resonant EP transport. The relevance of EP phase space modifications by instabilities is first discussed in terms of predicted fast ion distribution. Results are compared with those from a simple, ad-hoc diffusive model. It is then shown that the phase-space resolved model can also provide additional insight into important issues such as internal consistency of the simulations and mode stability through the analysis of the power exchanged between energetic particles and the instabilities.« less

Identification and control of plasma vertical position using neural network in Damavand tokamak.

PubMed

Rasouli, H; Rasouli, C; Koohi, A

2013-02-01

In this work, a nonlinear model is introduced to determine the vertical position of the plasma column in Damavand tokamak. Using this model as a simulator, a nonlinear neural network controller has been designed. In the first stage, the electronic drive and sensory circuits of Damavand tokamak are modified. These circuits can control the vertical position of the plasma column inside the vacuum vessel. Since the vertical position of plasma is an unstable parameter, a direct closed loop system identification algorithm is performed. In the second stage, a nonlinear model is identified for plasma vertical position, based on the multilayer perceptron (MLP) neural network (NN) structure. Estimation of simulator parameters has been performed by back-propagation error algorithm using Levenberg-Marquardt gradient descent optimization technique. The model is verified through simulation of the whole closed loop system using both simulator and actual plant in similar conditions. As the final stage, a MLP neural network controller is designed for simulator model. In the last step, online training is performed to tune the controller parameters. Simulation results justify using of the NN controller for the actual plant.

Long-wavelength microinstabilities in toroidal plasmas*

NASA Astrophysics Data System (ADS)

Tang, W. M.; Rewoldt, G.

1993-07-01

Realistic kinetic toroidal eigenmode calculations have been carried out to support a proper assessment of the influence of long-wavelength microturbulence on transport in tokamak plasmas. In order to efficiently evaluate large-scale kinetic behavior extending over many rational surfaces, significant improvements have been made to a toroidal finite element code used to analyze the fully two-dimensional (r,θ) mode structures of trapped-ion and toroidal ion temperature gradient (ITG) instabilities. It is found that even at very long wavelengths, these eigenmodes exhibit a strong ballooning character with the associated radial structure relatively insensitive to ion Landau damping at the rational surfaces. In contrast to the long-accepted picture that the radial extent of trapped-ion instabilities is characterized by the ion-gyroradius-scale associated with strong localization between adjacent rational surfaces, present results demonstrate that under realistic conditions, the actual scale is governed by the large-scale variations in the equilibrium gradients. Applications to recent measurements of fluctuation properties in Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. (International Atomic Energy Agency, Vienna, 1985), Vol. 1, p. 29] L-mode plasmas indicate that the theoretical trends appear consistent with spectral characteristics as well as rough heuristic estimates of the transport level. Benchmarking calculations in support of the development of a three-dimensional toroidal gyrokinetic code indicate reasonable agreement with respect to both the properties of the eigenfunctions and the magnitude of the eigenvalues during the linear phase of the simulations of toroidal ITG instabilities.

Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

NASA Astrophysics Data System (ADS)

Anderl, R. A.; Causey, R. A.; Davis, J. W.; Doerner, R. P.; Federici, G.; Haasz, A. A.; Longhurst, G. R.; Wampler, W. R.; Wilson, K. L.

Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions show a difference of several orders of magnitude in both inventory and permeation rate to the coolant.

A key to improved ion core confinement in the JET tokamak: ion stiffness mitigation due to combined plasma rotation and low magnetic shear.

PubMed

Mantica, P; Angioni, C; Challis, C; Colyer, G; Frassinetti, L; Hawkes, N; Johnson, T; Tsalas, M; deVries, P C; Weiland, J; Baiocchi, B; Beurskens, M N A; Figueiredo, A C A; Giroud, C; Hobirk, J; Joffrin, E; Lerche, E; Naulin, V; Peeters, A G; Salmi, A; Sozzi, C; Strintzi, D; Staebler, G; Tala, T; Van Eester, D; Versloot, T

2011-09-23

New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. [Phys. Rev. Lett. 102, 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future devices will require sufficient rotational shear and the capability of q profile manipulation.

A survey of electron Bernstein wave heating and current drive potential for spherical tokamaks

NASA Astrophysics Data System (ADS)

Urban, Jakub; Decker, Joan; Peysson, Yves; Preinhaelter, Josef; Shevchenko, Vladimir; Taylor, Gary; Vahala, Linda; Vahala, George

2011-08-01

The electron Bernstein wave (EBW) is typically the only wave in the electron cyclotron (EC) range that can be applied in spherical tokamaks for heating and current drive (H&CD). Spherical tokamaks (STs) operate generally in high-β regimes, in which the usual EC O- and X-modes are cut off. In this case, EBWs seem to be the only option that can provide features similar to the EC waves—controllable localized H&CD that can be used for core plasma heating as well as for accurate plasma stabilization. The EBW is a quasi-electrostatic wave that can be excited by mode conversion from a suitably launched O- or X-mode; its propagation further inside the plasma is strongly influenced by the plasma parameters. These rather awkward properties make its application somewhat more difficult. In this paper we perform an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX). Coupled ray-tracing (AMR) and Fokker-Planck (LUKE) codes are employed to simulate EBWs of varying frequencies and launch conditions, which are the fundamental EBW parameters that can be chosen and controlled. Our results indicate that an efficient and universal EBW H&CD system is indeed viable. In particular, power can be deposited and current reasonably efficiently driven across the whole plasma radius. Such a system could be controlled by a suitably chosen launching antenna vertical position and would also be sufficiently robust.

Modeling of 3D magnetic equilibrium effects on edge turbulence stability during RMP ELM suppression in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilcox, R. S.; Wingen, Andreas; Cianciosa, Mark R.

Some recent experimental observations have found turbulent fluctuation structures that are non-axisymmetric in a tokamak with applied 3D fields. Here, two fluid resistive effects are shown to produce changes relevant to turbulent transport in the modeled 3D magnetohydrodynamic (MHD) equilibrium of tokamak pedestals with these 3D fields applied. Ideal MHD models are insufficient to reproduce the relevant effects. By calculating the ideal 3D equilibrium using the VMEC code, the geometric shaping parameters that determine linear turbulence stability, including the normal curvature and local magnetic shear, are shown to be only weakly modified by applied 3D fields in the DIII-D tokamak.more » These ideal MHD effects are therefore not sufficient to explain the observed changes to fluctuations and transport. Using the M3D-C1 code to model the 3D equilibrium, density is shown to be redistributed on flux surfaces in the pedestal when resistive two fluid effects are included, while islands are screened by rotation in this region. Furthermore, the redistribution of density results in density and pressure gradient scale lengths that vary within pedestal flux surfaces between different helically localized flux tubes. This would produce different drive terms for trapped electron mode and kinetic ballooning mode turbulence, the latter of which is expected to be the limiting factor for pedestal pressure gradients in DIII-D.« less

Electron cyclotron emission from nonthermal tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harvey, R.W.; O'Brien, M.R.; Rozhdestvensky, V.V.

1993-02-01

Electron cyclotron emission can be a sensitive indicator of nonthermal electron distributions. A new, comprehensive ray-tracing and cyclotron emission code that is aimed at predicting and interpreting the cyclotron emission from tokamak plasmas is described. The radiation transfer equation is solved along Wentzel--Kramers--Brillouin (WKB) rays using a fully relativistic calculation of the emission and absorption from electron distributions that are gyrotropic and toroidally symmetric, but may be otherwise arbitrary functions of the constants of motion. Using a radial array of electron distributions obtained from a bounce-averaged Fokker--Planck code modeling dc electron field and electron cyclotron heating effects, the cyclotron emissionmore » spectra are obtained. A pronounced strong nonthermal cyclotron emission feature that occurs at frequencies relativistically downshifted to second harmonic cyclotron frequencies outside the tokamak is calculated, in agreement with experimental results from the DIII-D [J. L. Luxon and L. G. Davies, Fusion Technol. [bold 8], 441 (1985)] and FT-1 [D. G. Bulyginsky [ital et] [ital al]., in [ital Proceedings] [ital of] [ital the] 15[ital th] [ital European] [ital Conference] [ital on] [ital Controlled] [ital Fusion] [ital and] [ital Plasma] [ital Heating], Dubrovnik, 1988 (European Physical Society, Petit-Lancy, 1988), Vol. 12B, Part II, p. 823] tokamaks. The calculations indicate the presence of a strong loss mechanism that operates on electrons in the 100--150 keV energy range.« less

Modeling of 3D magnetic equilibrium effects on edge turbulence stability during RMP ELM suppression in tokamaks

DOE PAGES

Wilcox, R. S.; Wingen, Andreas; Cianciosa, Mark R.; ...

2017-07-28

Some recent experimental observations have found turbulent fluctuation structures that are non-axisymmetric in a tokamak with applied 3D fields. Here, two fluid resistive effects are shown to produce changes relevant to turbulent transport in the modeled 3D magnetohydrodynamic (MHD) equilibrium of tokamak pedestals with these 3D fields applied. Ideal MHD models are insufficient to reproduce the relevant effects. By calculating the ideal 3D equilibrium using the VMEC code, the geometric shaping parameters that determine linear turbulence stability, including the normal curvature and local magnetic shear, are shown to be only weakly modified by applied 3D fields in the DIII-D tokamak.more » These ideal MHD effects are therefore not sufficient to explain the observed changes to fluctuations and transport. Using the M3D-C1 code to model the 3D equilibrium, density is shown to be redistributed on flux surfaces in the pedestal when resistive two fluid effects are included, while islands are screened by rotation in this region. Furthermore, the redistribution of density results in density and pressure gradient scale lengths that vary within pedestal flux surfaces between different helically localized flux tubes. This would produce different drive terms for trapped electron mode and kinetic ballooning mode turbulence, the latter of which is expected to be the limiting factor for pedestal pressure gradients in DIII-D.« less

Total fluid pressure imbalance in the scrape-off layer of tokamak plasmas

NASA Astrophysics Data System (ADS)

Churchill, R. M.; Canik, J. M.; Chang, C. S.; Hager, R.; Leonard, A. W.; Maingi, R.; Nazikian, R.; Stotler, D. P.

2017-04-01

Simulations using the fully kinetic neoclassical code XGCa (X-point included guiding- center axisymmetric) were undertaken to explore the impact of kinetic effects on scrape-off layer (SOL) physics in DIII-D H-mode plasmas. XGCa is a total-f, gyrokinetic code which self-consistently calculates the axisymmetric electrostatic potential and plasma dynamics, and includes modules for Monte Carlo neutral transport. Previously presented XGCa results showed several noteworthy features, including large variations of ion density and pressure along field lines in the SOL, experimentally relevant levels of SOL parallel ion flow (Mach number  ˜ 0.5), skewed ion distributions near the sheath entrance leading to subsonic flow there, and elevated sheath potentials (Churchill 2016 Nucl. Mater. Energy 1-6). In this paper, we explore in detail the question of pressure balance in the SOL, as it was observed in the simulation that there was a large deviation from a simple total pressure balance (the sum of ion and electron static pressure plus ion inertia). It will be shown that both the contributions from the ion viscosity (driven by ion temperature anisotropy) and neutral source terms can be substantial, and should be retained in the parallel momentum equation in the SOL, but still falls short of accounting for the observed fluid pressure imbalance in the XGCa simulation results.

Monte-Carlo Orbit/Full Wave Simulation of Fast Alfvén Wave (FW) Damping on Resonant Ions in Tokamaks

NASA Astrophysics Data System (ADS)

Choi, M.; Chan, V. S.; Tang, V.; Bonoli, P.; Pinsker, R. I.; Wright, J.

2005-09-01

To simulate the resonant interaction of fast Alfvén wave (FW) heating and Coulomb collisions on energetic ions, including finite orbit effects, a Monte-Carlo code ORBIT-RF has been coupled with a 2D full wave code TORIC4. ORBIT-RF solves Hamiltonian guiding center drift equations to follow trajectories of test ions in 2D axisymmetric numerical magnetic equilibrium under Coulomb collisions and ion cyclotron radio frequency quasi-linear heating. Monte-Carlo operators for pitch-angle scattering and drag calculate the changes of test ions in velocity and pitch angle due to Coulomb collisions. A rf-induced random walk model describing fast ion stochastic interaction with FW reproduces quasi-linear diffusion in velocity space. FW fields and its wave numbers from TORIC are passed on to ORBIT-RF to calculate perpendicular rf kicks of resonant ions valid for arbitrary cyclotron harmonics. ORBIT-RF coupled with TORIC using a single dominant toroidal and poloidal wave number has demonstrated consistency of simulations with recent DIII-D FW experimental results for interaction between injected neutral-beam ions and FW, including measured neutron enhancement and enhanced high energy tail. Comparison with C-Mod fundamental heating discharges also yielded reasonable agreement.

Comparing simulation of plasma turbulence with experiment. II. Gyrokinetic simulations

NASA Astrophysics Data System (ADS)

Ross, David W.; Dorland, William

2002-12-01

The direct quantitative correspondence between theoretical predictions and the measured plasma fluctuations and transport is tested by performing nonlinear gyrokinetic simulations with the GS2 code. This is a continuation of previous work with gyrofluid simulations [D. W. Ross et al., Phys. Plasmas 9, 177 (2002)], and the same L-mode reference discharge in the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)] is studied. The simulated turbulence is dominated by ion temperature gradient (ITG) modes, corrected by trapped-electron, passing-electron and impurity effects. The energy fluxes obtained in the gyrokinetic simulations are comparable to, even somewhat higher than, those of the earlier work, and the simulated ion thermal transport, corrected for E×B flow shear, exceeds the experimental value by more than a factor of 2. The simulation also overestimates the density fluctuation level. Varying the local temperature gradient shows a stiff response in the flux and an apparent up-shift from the linear mode threshold [A. M. Dimits et al., Phys. Plasmas 7, 969 (2000)]. This effect is insufficient, within the estimated error, to bring the results into conformity with the experiment.

Active spectroscopic measurements of the bulk deuterium properties in the DIII-D tokamak (invited).

PubMed

Grierson, B A; Burrell, K H; Chrystal, C; Groebner, R J; Kaplan, D H; Heidbrink, W W; Muñoz Burgos, J M; Pablant, N A; Solomon, W M; Van Zeeland, M A

2012-10-01

The neutral-beam induced D(α) emission spectrum contains a wealth of information such as deuterium ion temperature, toroidal rotation, density, beam emission intensity, beam neutral density, and local magnetic field strength magnitude |B| from the Stark-split beam emission spectrum, and fast-ion D(α) emission (FIDA) proportional to the beam-injected fast ion density. A comprehensive spectral fitting routine which accounts for all photoemission processes is employed for the spectral analysis. Interpretation of the measurements to determine physically relevant plasma parameters is assisted by the use of an optimized viewing geometry and forward modeling of the emission spectra using a Monte-Carlo 3D simulation code.

Toroidal gyrofluid equations for simulations of tokamak turbulence

NASA Astrophysics Data System (ADS)

Beer, M. A.; Hammett, G. W.

1996-11-01

A set of nonlinear gyrofluid equations for simulations of tokamak turbulence are derived by taking moments of the nonlinear toroidal gyrokinetic equation. The moment hierarchy is closed with approximations that model the kinetic effects of parallel Landau damping, toroidal drift resonances, and finite Larmor radius effects. These equations generalize the work of Dorland and Hammett [Phys. Fluids B 5, 812 (1993)] to toroidal geometry by including essential toroidal effects. The closures for phase mixing from toroidal ∇B and curvature drifts take the basic form presented in Waltz et al. [Phys. Fluids B 4, 3138 (1992)], but here a more rigorous procedure is used, including an extension to higher moments, which provides significantly improved accuracy. In addition, trapped ion effects and collisions are incorporated. This reduced set of nonlinear equations accurately models most of the physics considered important for ion dynamics in core tokamak turbulence, and is simple enough to be used in high resolution direct numerical simulations.

Stability analysis of ELMs in long-pulse discharges with ELITE code on EAST tokamak

NASA Astrophysics Data System (ADS)

Wang, Y. F.; Xu, G. S.; Wan, B. N.; Li, G. Q.; Yan, N.; Li, Y. L.; Wang, H. Q.; Peng, Y.-K. Martin; Xia, T. Y.; Ding, S. Y.; Chen, R.; Yang, Q. Q.; Liu, H. Q.; Zang, Q.; Zhang, T.; Lyu, B.; Xu, J. C.; Feng, W.; Wang, L.; Chen, Y. J.; Luo, Z. P.; Hu, G. H.; Zhang, W.; Shao, L. M.; Ye, Y.; Lan, H.; Chen, L.; Li, J.; Zhao, N.; Wang, Q.; Snyder, P. B.; Liang, Y.; Qian, J. P.; Gong, X. Z.; EAST team

2018-05-01

One challenge in long-pulse and high performance tokamak operation is to control the edge localized modes (ELMs) to reduce the transient heat load on plasma facing components. Minute-scale discharges in H-mode have been achieved repeatedly on Experimental Advanced Superconducting Tokamak (EAST) since the 2016 campaign and understanding the characteristics of the ELMs in these discharges can be helpful for effective ELM control in long-pulse discharges. The kinetic profile diagnostics recently developed on EAST make it possible to perform the pedestal stability analysis quantitatively. Pedestal stability calculation of a typical long-pulse discharge with ELITE code is presented. The ideal linear stability results show that the ELM is dominated by toroidal mode number n around 10–15 and the most unstable mode structure is mainly localized in the steep pressure gradient region, which is consistent with experimental results. Compared with a typical type-I ELM discharge with larger total plasma current (I p = 600 kA), pedestal in the long-pulse H-mode discharge (I p = 450 kA) is more stable in peeling-ballooning instability and its critical peak pressure gradient is evaluated to be 65% of the former. Two important features of EAST tokamak in the long-pulse discharge are presented by comparison with other tokamaks, including a wider pedestal correlated with the poloidal pedestal beta and a smaller inverse aspect ratio and their effects on the pedestal stability are discussed. The effects of uncertainties in measurements on the linear stability results are also analyzed, including the edge electron density profile position, the separatrix position and the line-averaged effective ion charge {Z}{{e}{{f}}{{f}}} value.

An atomic and molecular fluid model for efficient edge-plasma transport simulations at high densities

NASA Astrophysics Data System (ADS)

Rognlien, Thomas; Rensink, Marvin

2016-10-01

Transport simulations for the edge plasma of tokamaks and other magnetic fusion devices requires the coupling of plasma and recycling or injected neutral gas. There are various neutral models used for this purpose, e.g., atomic fluid model, a Monte Carlo particle models, transition/escape probability methods, and semi-analytic models. While the Monte Carlo method is generally viewed as the most accurate, it is time consuming, which becomes even more demanding for device simulations of high densities and size typical of fusion power plants because the neutral collisional mean-free path becomes very small. Here we examine the behavior of an extended fluid neutral model for hydrogen that includes both atoms and molecules, which easily includes nonlinear neutral-neutral collision effects. In addition to the strong charge-exchange between hydrogen atoms and ions, elastic scattering is included among all species. Comparisons are made with the DEGAS 2 Monte Carlo code. Work performed for U.S. DoE by LLNL under Contract DE-AC52-07NA27344.

Study of neutron generation in the compact tokamak TUMAN-3M in support of a tokamak-based fusion neutron source

NASA Astrophysics Data System (ADS)

Kornev, V. A.; Askinazi, L. G.; Belokurov, A. A.; Chernyshev, F. V.; Lebedev, S. V.; Melnik, A. D.; Shabelsky, A. A.; Tukachinsky, A. S.; Zhubr, N. A.

2017-12-01

The paper presents DD neutron flux measurements in neutron beam injection (NBI) experiments aimed at the optimization of target plasma and heating beam parameters to achieve maximum neutron flux in the TUMAN-3M compact tokamak. Two ion sources of different design were used, which allowed the separation of the beam’s energy and power influence on the neutron rate. Using the database of experiments performed with the two ion sources, an empirical scaling was derived describing the neutron rate dependence on the target plasma and heating beam parameters. Numerical modeling of the neutron rate in the NBI experiments performed using the ASTRA transport code showed good agreement with the scaling.

Simulation of wave interactions with MHD

NASA Astrophysics Data System (ADS)

Batchelor, D.; Alba, C.; Bateman, G.; Bernholdt, D.; Berry, L.; Bonoli, P.; Bramley, R.; Breslau, J.; Chance, M.; Chen, J.; Choi, M.; Elwasif, W.; Fu, G.; Harvey, R.; Jaeger, E.; Jardin, S.; Jenkins, T.; Keyes, D.; Klasky, S.; Kruger, S.; Ku, L.; Lynch, V.; McCune, D.; Ramos, J.; Schissel, D.; Schnack, D.; Wright, J.

2008-07-01

The broad scientific objectives of the SWIM (Simulation 01 Wave Interaction with MHD) project are twofold: (1) improve our understanding of interactions that both radio frequency (RF) wave and particle sources have on extended-MHD phenomena, and to substantially improve our capability for predicting and optimizing the performance of burning plasmas in devices such as ITER: and (2) develop an integrated computational system for treating multiphysics phenomena with the required flexibility and extensibility to serve as a prototype for the Fusion Simulation Project. The Integrated Plasma Simulator (IPS) has been implemented. Presented here are initial physics results on RP effects on MHD instabilities in tokamaks as well as simulation results for tokamak discharge evolution using the IPS.

Alpha-channeling simulation experiment in the DIII-D tokamak.

PubMed

Wong, K L; Budny, R; Nazikian, R; Petty, C C; Greenfield, C M; Heidbrink, W W; Ruskov, E

2004-08-20

Alfvén instabilities can reduce the central magnetic shear via redistribution of energetic ions. They can sustain a steady state internal transport barrier as demonstrated in this DIII-D tokamak experiment. Improvement in burning plasma performance based on this mechanism is discussed.

Nonlinear Simulation of DIII-D Plasma and Poloidal Systems Using DINA and Simulink

NASA Astrophysics Data System (ADS)

Walker, M. L.; Leuer, J. A.; Deranian, R. D.; Humphreys, D. A.; Khayrutdinov, R. R.

2002-11-01

Hardware-in-the-loop simulation capability was developed previously for poloidal shape control testing using Matlab Simulink [1]. This has been upgraded by replacing a linearized plasma model with the DINA nonlinear plasma evolution code [2]. In addition to its use for shape control studies, this new capability will allow study of current profile control using the DINA model of electron cyclotron current drive (ECCD) and current profile information soon to be available from the Plasma Control System (PCS) real time EFIT [3] calculation. We describe the incorporation of DINA into the Simulink DIII-D tokamak systems model and results of validating this combined model against DIII-D data. \\vspace0.1em [1] J.A. Leuer, et al., 18th IEEE/NPSS SOFE (1999), p. 531. [2] R.R. Khayrutdinov, V.E. Lukash, J. Comput. Phys. 109, 193 (1993). [3] J.R. Ferron, et al., Nucl. Fusion 38, 1055 (1988).

Transport Barriers in Bootstrap Driven Tokamaks

NASA Astrophysics Data System (ADS)

Staebler, Gary

2017-10-01

Maximizing the bootstrap current in a tokamak, so that it drives a high fraction of the total current, reduces the external power required to drive current by other means. Improved energy confinement, relative to empirical scaling laws, enables a reactor to more fully take advantage of the bootstrap driven tokamak. Experiments have demonstrated improved energy confinement due to the spontaneous formation of an internal transport barrier in high bootstrap fraction discharges. Gyrokinetic analysis, and quasilinear predictive modeling, demonstrates that the observed transport barrier is due to the suppression of turbulence primarily due to the large Shafranov shift. ExB velocity shear does not play a significant role in the transport barrier due to the high safety factor. It will be shown, that the Shafranov shift can produce a bifurcation to improved confinement in regions of positive magnetic shear or a continuous reduction in transport for weak or negative magnetic shear. Operation at high safety factor lowers the pressure gradient threshold for the Shafranov shift driven barrier formation. The ion energy transport is reduced to neoclassical and electron energy and particle transport is reduced, but still turbulent, within the barrier. Deeper into the plasma, very large levels of electron transport are observed. The observed electron temperature profile is shown to be close to the threshold for the electron temperature gradient (ETG) mode. A large ETG driven energy transport is qualitatively consistent with recent multi-scale gyrokinetic simulations showing that reducing the ion scale turbulence can lead to large increase in the electron scale transport. A new saturation model for the quasilinear TGLF transport code, that fits these multi-scale gyrokinetic simulations, can match the data if the impact of zonal flow mixing on the ETG modes is reduced at high safety factor. This work was supported by the U.S. Department of Energy under DE-FG02-95ER54309 and DE-FC02-04ER54698.

Analysis of Radiation Transport Due to Activated Coolant in the ITER Neutral Beam Injection Cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Royston, Katherine; Wilson, Stephen C.; Risner, Joel M.

Detailed spatial distributions of the biological dose rate due to a variety of sources are required for the design of the ITER tokamak facility to ensure that all radiological zoning limits are met. During operation, water in the Integrated loop of Blanket, Edge-localized mode and vertical stabilization coils, and Divertor (IBED) cooling system will be activated by plasma neutrons and will flow out of the bioshield through a complex system of pipes and heat exchangers. This paper discusses the methods used to characterize the biological dose rate outside the tokamak complex due to 16N gamma radiation emitted by the activatedmore » coolant in the Neutral Beam Injection (NBI) cell of the tokamak building. Activated coolant will enter the NBI cell through the IBED Primary Heat Transfer System (PHTS), and the NBI PHTS will also become activated due to radiation streaming through the NBI system. To properly characterize these gamma sources, the production of 16N, the decay of 16N, and the flow of activated water through the coolant loops were modeled. The impact of conservative approximations on the solution was also examined. Once the source due to activated coolant was calculated, the resulting biological dose rate outside the north wall of the NBI cell was determined through the use of sophisticated variance reduction techniques. The AutomateD VAriaNce reducTion Generator (ADVANTG) software implements methods developed specifically to provide highly effective variance reduction for complex radiation transport simulations such as those encountered with ITER. Using ADVANTG with the Monte Carlo N-particle (MCNP) radiation transport code, radiation responses were calculated on a fine spatial mesh with a high degree of statistical accuracy. In conclusion, advanced visualization tools were also developed and used to determine pipe cell connectivity, to facilitate model checking, and to post-process the transport simulation results.« less

Analysis of Radiation Transport Due to Activated Coolant in the ITER Neutral Beam Injection Cell

DOE PAGES

Royston, Katherine; Wilson, Stephen C.; Risner, Joel M.; ...

2017-07-26

Detailed spatial distributions of the biological dose rate due to a variety of sources are required for the design of the ITER tokamak facility to ensure that all radiological zoning limits are met. During operation, water in the Integrated loop of Blanket, Edge-localized mode and vertical stabilization coils, and Divertor (IBED) cooling system will be activated by plasma neutrons and will flow out of the bioshield through a complex system of pipes and heat exchangers. This paper discusses the methods used to characterize the biological dose rate outside the tokamak complex due to 16N gamma radiation emitted by the activatedmore » coolant in the Neutral Beam Injection (NBI) cell of the tokamak building. Activated coolant will enter the NBI cell through the IBED Primary Heat Transfer System (PHTS), and the NBI PHTS will also become activated due to radiation streaming through the NBI system. To properly characterize these gamma sources, the production of 16N, the decay of 16N, and the flow of activated water through the coolant loops were modeled. The impact of conservative approximations on the solution was also examined. Once the source due to activated coolant was calculated, the resulting biological dose rate outside the north wall of the NBI cell was determined through the use of sophisticated variance reduction techniques. The AutomateD VAriaNce reducTion Generator (ADVANTG) software implements methods developed specifically to provide highly effective variance reduction for complex radiation transport simulations such as those encountered with ITER. Using ADVANTG with the Monte Carlo N-particle (MCNP) radiation transport code, radiation responses were calculated on a fine spatial mesh with a high degree of statistical accuracy. In conclusion, advanced visualization tools were also developed and used to determine pipe cell connectivity, to facilitate model checking, and to post-process the transport simulation results.« less

A Finite-Orbit-Width Fokker-Planck solver for modeling of RF Current Drive in ITER

NASA Astrophysics Data System (ADS)

Petrov, Yu. V.; Harvey, R. W.

2017-10-01

The bounce-average (BA) finite-difference Fokker-Planck (FP) code CQL3D now includes the essential physics to describe the RF heating of Finite-Orbit-Width (FOW) ions in tokamaks. The FP equation is reformulated in terms of constants-of-motion coordinates, which we select to be particle speed, pitch angle, and major radius on the equatorial plane thus obtaining the distribution function directly at this location. A recent development is the capability to obtain solution simultaneously for FOW ions and Zero-Orbit-Width (ZOW) electrons. As a practical application, the code is used for simulation of alpha-particle heating by high-harmonic waves in ITER scenarios. Coupling of high harmonic or helicon fast waves power to electrons is a promising current drive (CD) scenario for high beta plasmas. However, the efficiency of current drive can be diminished by parasitic channeling of RF power into fast ions such as alphas or NBI-produced deuterons, through finite Larmor-radius effects. Based on simulations, we formulate conditions where the fast ions absorb less than 10% of RF power. Supported by USDOE Grants ER54649, ER54744, and SC0006614.

Implementation of the 3D edge plasma code EMC3-EIRENE on NSTX

DOE PAGES

Lore, J. D.; Canik, J. M.; Feng, Y.; ...

2012-05-09

The 3D edge transport code EMC3-EIRENE has been applied for the first time to the NSTX spherical tokamak. A new disconnected double null grid has been developed to allow the simulation of plasma where the radial separation of the inner and outer separatrix is less than characteristic widths (e.g. heat flux width) at the midplane. Modelling results are presented for both an axisymmetric case and a case where 3D magnetic field is applied in an n = 3 configuration. In the vacuum approximation, the perturbed field consists of a wide region of destroyed flux surfaces and helical lobes which aremore » a mixture of long and short connection length field lines formed by the separatrix manifolds. This structure is reflected in coupled 3D plasma fluid (EMC3) and kinetic neutral particle (EIRENE) simulations. The helical lobes extending inside of the unperturbed separatrix are filled in by hot plasma from the core. The intersection of the lobes with the divertor results in a striated flux footprint pattern on the target plates. As a result, profiles of divertor heat and particle fluxes are compared with experimental data, and possible sources of discrepancy are discussed.« less

Propagation of radio frequency waves through density fluctuations

NASA Astrophysics Data System (ADS)

Valvis, S. I.; Papagiannis, P.; Papadopoulos, A.; Hizanidis, K.; Glytsis, E.; Bairaktaris, F.; Zisis, A.; Tigelis, I.; Ram, A. K.

2017-10-01

On their way to the core of a tokamak plasma, radio frequency (RF) waves, excited in the vacuum region, have to propagate through a variety of density fluctuations in the edge region. These fluctuations include coherent structures, like blobs that can be field aligned or not, as well as turbulent and filamentary structures. We have been studying the effect of fluctuations on RF propagation using both theoretical (analytical) and computational models. The theoretical results are being compared with those obtained by two different numerical codes ``a Finite Difference Frequency Domain code and the commercial COMSOL package. For plasmas with arbitrary distribution of coherent and turbulent fluctuations, we have formulated an effective dielectric permittivity of the edge plasma. This permittivity tensor is then used in numerical simulations to study the effect of multi-scale turbulence on RF waves. We not only consider plane waves but also Gaussian beams in the electron cyclotron and lower hybrid range of frequencies. The analytical theory and results from simulations on the propagation of RF waves will be presented. Supported in part by the Hellenic National Programme on Controlled Thermonuclear Fusion associated with the EUROfusion Consortium and by DoE Grant DE-FG02-91ER-54109.

Modeling of induction-linac based free-electron laser amplifiers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jong, R.A.; Fawley, W.M.; Scharlemann, E.T.

We describe the modeling of an induction-linac based free-electron laser (IFEL) amplifier for producing multimegawatt levels of microwave power. We have used the Lawrence Livermore National Laboratory (LLNL) free-electron laser simulation code, FRED, and the simulation code for sideband calculations, GINGER for this study. For IFEL amplifiers in the frequency range of interest (200 to 600 GHz), we have devised a wiggler design strategy which incorporates a tapering algorithm that is suitable for free-electron laser (FEL) systems with moderate space-charge effects and that minimizes spontaneous noise growth at frequencies below the fundamental, while enhancing the growth of the signal atmore » the fundamental. In addition, engineering design considerations of the waveguide wall loading and electron beam fill factor in the waveguide set limits on the waveguide dimensions, the wiggler magnet gap spacing, the wiggler period, and the minimum magnetic field strength in the tapered region of the wiggler. As an example, we shall describe an FEL amplifier designed to produce an average power of about 10 MW at a frequency of 280 GHz to be used for electron cyclotron resonance heating of tokamak fusion devices. 17 refs., 4 figs.« less

AORSA full wave calculations of helicon waves in DIII-D and ITER

NASA Astrophysics Data System (ADS)

Lau, C.; Jaeger, E. F.; Bertelli, N.; Berry, L. A.; Green, D. L.; Murakami, M.; Park, J. M.; Pinsker, R. I.; Prater, R.

2018-06-01

Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D, FNSF, and DEMO tokamaks. Previous ray tracing modeling using GENRAY predicts strong single pass absorption and current drive in the mid-radius region on DIII-D in high beta tokamak discharges. The full wave code AORSA, which is valid to all order of Larmor radius and can resolve arbitrary ion cyclotron harmonics, has been used to validate the ray tracing technique. If the scrape-off-layer (SOL) is ignored in the modeling, AORSA agrees with GENRAY in both the amplitude and location of driven current for DIII-D and ITER cases. These models also show that helicon current drive can possibly be an efficient current drive actuator for ITER. Previous GENRAY analysis did not include the SOL. AORSA has also been used to extend the simulations to include the SOL and to estimate possible power losses of helicon waves in the SOL. AORSA calculations show that another mode can propagate in the SOL and lead to significant (~10%–20%) SOL losses at high SOL densities. Optimizing the SOL density profile can reduce these SOL losses to a few percent.

AORSA full wave calculations of helicon waves in DIII-D and ITER

DOE PAGES

Lau, Cornwall; Jaeger, E.F.; Bertelli, Nicola; ...

2018-04-11

Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D, FNSF, and DEMO tokamaks. Previous ray tracing modeling using GENRAY predicts strong single pass absorption and current drive in the mid-radius region on DIII-D in high beta tokamak discharges. The full wave code AORSA, which is valid to all order of Larmor radius and can resolve arbitrary ion cyclotron harmonics, has been used to validate the ray tracing technique. If the scrape-off-layer (SOL) is ignored in the modeling, AORSA agrees with GENRAY in both the amplitude and location of driven current for DIII-D and ITER cases.more » These models also show that helicon current drive can possibly be an efficient current drive actuator for ITER. Previous GENRAY analysis did not include the SOL. AORSA has also been used to extend the simulations to include the SOL and to estimate possible power losses of helicon waves in the SOL. AORSA calculations show that another mode can propagate in the SOL and lead to significant (~10-20%) SOL losses at high SOL densities. Optimizing the SOL density profile can reduce these SOL losses to a few percent.« less

AORSA full wave calculations of helicon waves in DIII-D and ITER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lau, Cornwall; Jaeger, E.F.; Bertelli, Nicola

Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D, FNSF, and DEMO tokamaks. Previous ray tracing modeling using GENRAY predicts strong single pass absorption and current drive in the mid-radius region on DIII-D in high beta tokamak discharges. The full wave code AORSA, which is valid to all order of Larmor radius and can resolve arbitrary ion cyclotron harmonics, has been used to validate the ray tracing technique. If the scrape-off-layer (SOL) is ignored in the modeling, AORSA agrees with GENRAY in both the amplitude and location of driven current for DIII-D and ITER cases.more » These models also show that helicon current drive can possibly be an efficient current drive actuator for ITER. Previous GENRAY analysis did not include the SOL. AORSA has also been used to extend the simulations to include the SOL and to estimate possible power losses of helicon waves in the SOL. AORSA calculations show that another mode can propagate in the SOL and lead to significant (~10-20%) SOL losses at high SOL densities. Optimizing the SOL density profile can reduce these SOL losses to a few percent.« less

A fluid modeling perspective on the tokamak power scrape-off width using SOLPS-ITER

NASA Astrophysics Data System (ADS)

Meier, Eric

2016-10-01

SOLPS-ITER, a 2D fluid code, is used to conduct the first fluid modeling study of the physics behind the power scrape-off width (λq). When drift physics are activated in the code, λq is insensitive to changes in toroidal magnetic field (Bt), as predicted by the 0D heuristic drift (HD) model developed by Goldston. Using the HD model, which quantitatively agrees with regression analysis of a multi-tokamak database, λq in ITER is projected to be 1 mm instead of the previously assumed 4 mm, magnifying the challenge of maintaining the peak divertor target heat flux below the technological limit. These simulations, which use DIII-D H-mode experimental conditions as input, and reproduce the observed high-recycling, attached outer target plasma, allow insights into the scrape-off layer (SOL) physics that set λq. Independence of λq with respect to Bt suggests that SOLPS-ITER captures basic HD physics: the effect of Bt on the particle dwell time ( Bt) cancels with the effect on drift speed ( 1 /Bt), fixing the SOL plasma density width, and dictating λq. Scaling with plasma current (Ip), however, is much weaker than the roughly 1 /Ip dependence predicted by the HD model. Simulated net cross-separatrix particle flux due to magnetic drifts exceeds the anomalous particle transport, and a Pfirsch-Schluter-like SOL flow pattern is established. Up-down ion pressure asymmetry enables the net magnetic drift flux. Drifts establish in-out temperature asymmetry, and an associated thermoelectric current carries significant heat flux to the outer target. The density fall-off length in the SOL is similar to the electron temperature fall-off length, as observed experimentally. Finally, opportunities and challenges foreseen in ongoing work to extrapolate SOLPS-ITER and the HD model to ITER and future machines will be discussed. Supported by U.S. Department of Energy Contract DESC0010434.

A post-processor for the PEST code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Priesche, S.; Manickam, J.; Johnson, J.L.

1992-01-01

A new post-processor has been developed for use with output from the PEST tokamak stability code. It allows us to use quantities calculated by PEST and take better advantage of the physical picture of the plasma instability which they can provide. This will improve comparison with experimentally measured quantities as well as facilitate understanding of theoretical studies.

Numerical simulation of plasma response to externally applied resonant magnetic perturbation on the J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Bicheng, LI; Zhonghe, JIANG; Jian, LV; Xiang, LI; Bo, RAO; Yonghua, DING

2018-05-01

Nonlinear magnetohydrodynamic (MHD) simulations of an equilibrium on the J-TEXT tokamak with applied resonant magnetic perturbations (RMPs) are performed with NIMROD (non-ideal MHD with rotation, open discussion). Numerical simulation of plasma response to RMPs has been developed to investigate magnetic topology, plasma density and rotation profile. The results indicate that the pure applied RMPs can stimulate 2/1 mode as well as 3/1 mode by the toroidal mode coupling, and finally change density profile by particle transport. At the same time, plasma rotation plays an important role during the entire evolution process.

Modelling of the EAST lower-hybrid current drive experiment using GENRAY/CQL3D and TORLH/CQL3D

NASA Astrophysics Data System (ADS)

Yang, C.; Bonoli, P. T.; Wright, J. C.; Ding, B. J.; Parker, R.; Shiraiwa, S.; Li, M. H.

2014-12-01

The coupled GENRAY-CQL3D code has been used to do systematic ray-tracing and Fokker-Planck analysis for EAST Lower Hybrid wave Current Drive (LHCD) experiments. Despite being in the weak absorption regime, the experimental level of LH current drive is successfully simulated, by taking into account the variations in the parallel wavenumber due to the toroidal effect. The effect of radial transport of the fast LH electrons in EAST has also been studied, which shows that a modest amount of radial transport diffusion can redistribute the fast LH current significantly. Taking advantage of the new capability in GENRAY, the actual Scrape Off Layer (SOL) model with magnetic field, density, temperature, and geometry is included in the simulation for both the lower and the higher density cases, so that the collisional losses of Lower Hybrid Wave (LHW) power in the SOL has been accounted for, which together with fast electron losses can reproduce the LHCD experimental observations in different discharges of EAST. We have also analyzed EAST discharges where there is a significant ohmic contribution to the total current, and good agreement with experiment in terms of total current has been obtained. Also, the full-wave code TORLH has been used for the simulation of the LH physics in the EAST, including full-wave effects such as diffraction and focusing which may also play an important role in bridging the spectral gap. The comparisons between the GENRAY and the TORLH codes are done for both the Maxwellian and the quasi-linear electron Landau damping cases. These simulations represent an important addition to the validation studies of the GENRAY-CQL3D and TORLH models being used in weak absorption scenarios of tokamaks with large aspect ratio.

Interaction between neoclassical effects and ion temperature gradient turbulence in gradient- and flux-driven gyrokinetic simulations

NASA Astrophysics Data System (ADS)

Oberparleiter, M.; Jenko, F.; Told, D.; Doerk, H.; Görler, T.

2016-04-01

Neoclassical and turbulent transport in tokamaks has been studied extensively over the past decades, but their possible interaction remains largely an open question. The two are only truly independent if the length scales governing each of them are sufficiently separate, i.e., if the ratio ρ* between ion gyroradius and the pressure gradient scale length is small. This is not the case in particularly interesting regions such as transport barriers. Global simulations of a collisional ion-temperature-gradient-driven microturbulence performed with the nonlinear global gyrokinetic code Gene are presented. In particular, comparisons are made between systems with and without neoclassical effects. In fixed-gradient simulations, the modified radial electric field is shown to alter the zonal flow pattern such that a significant increase in turbulent transport is observed for ρ*≳1 /300 . Furthermore, the dependency of the flux on the collisionality changes. In simulations with fixed power input, we find that the presence of neoclassical effects decreases the frequency and amplitude of intermittent turbulent transport bursts (avalanches) and thus plays an important role for the self-organisation behaviour.

Simulation of the hybrid and steady state advanced operating modes in ITER

NASA Astrophysics Data System (ADS)

Kessel, C. E.; Giruzzi, G.; Sips, A. C. C.; Budny, R. V.; Artaud, J. F.; Basiuk, V.; Imbeaux, F.; Joffrin, E.; Schneider, M.; Murakami, M.; Luce, T.; St. John, Holger; Oikawa, T.; Hayashi, N.; Takizuka, T.; Ozeki, T.; Na, Y.-S.; Park, J. M.; Garcia, J.; Tucillo, A. A.

2007-09-01

Integrated simulations are performed to establish a physics basis, in conjunction with present tokamak experiments, for the operating modes in the International Thermonuclear Experimental Reactor (ITER). Simulations of the hybrid mode are done using both fixed and free-boundary 1.5D transport evolution codes including CRONOS, ONETWO, TSC/TRANSP, TOPICS and ASTRA. The hybrid operating mode is simulated using the GLF23 and CDBM05 energy transport models. The injected powers are limited to the negative ion neutral beam, ion cyclotron and electron cyclotron heating systems. Several plasma parameters and source parameters are specified for the hybrid cases to provide a comparison of 1.5D core transport modelling assumptions, source physics modelling assumptions, as well as numerous peripheral physics modelling. Initial results indicate that very strict guidelines will need to be imposed on the application of GLF23, for example, to make useful comparisons. Some of the variations among the simulations are due to source models which vary widely among the codes used. In addition, there are a number of peripheral physics models that should be examined, some of which include fusion power production, bootstrap current, treatment of fast particles and treatment of impurities. The hybrid simulations project to fusion gains of 5.6-8.3, βN values of 2.1-2.6 and fusion powers ranging from 350 to 500 MW, under the assumptions outlined in section 3. Simulations of the steady state operating mode are done with the same 1.5D transport evolution codes cited above, except the ASTRA code. In these cases the energy transport model is more difficult to prescribe, so that energy confinement models will range from theory based to empirically based. The injected powers include the same sources as used for the hybrid with the possible addition of lower hybrid. The simulations of the steady state mode project to fusion gains of 3.5-7, βN values of 2.3-3.0 and fusion powers of 290 to 415 MW, under the assumptions described in section 4. These simulations will be presented and compared with particular focus on the resulting temperature profiles, source profiles and peripheral physics profiles. The steady state simulations are at an early stage and are focused on developing a range of safety factor profiles with 100% non-inductive current.

Optimization of 3D Field Design

NASA Astrophysics Data System (ADS)

Logan, Nikolas; Zhu, Caoxiang

2017-10-01

Recent progress in 3D tokamak modeling is now leveraged to create a conceptual design of new external 3D field coils for the DIII-D tokamak. Using the IPEC dominant mode as a target spectrum, the Finding Optimized Coils Using Space-curves (FOCUS) code optimizes the currents and 3D geometry of multiple coils to maximize the total set's resonant coupling. The optimized coils are individually distorted in space, creating toroidal ``arrays'' containing a variety of shapes that often wrap around a significant poloidal extent of the machine. The generalized perturbed equilibrium code (GPEC) is used to determine optimally efficient spectra for driving total, core, and edge neoclassical toroidal viscosity (NTV) torque and these too provide targets for the optimization of 3D coil designs. These conceptual designs represent a fundamentally new approach to 3D coil design for tokamaks targeting desired plasma physics phenomena. Optimized coil sets based on plasma response theory will be relevant to designs for future reactors or on any active machine. External coils, in particular, must be optimized for reliable and efficient fusion reactor designs. Work supported by the US Department of Energy under DE-AC02-09CH11466.

Electron Cyclotron Current Drive Efficiency in General Tokamak Geometry and Its Application to Advanced Tokamak Plasmas

NASA Astrophysics Data System (ADS)

Lin-Liu, Y. R.; Chan, V. S.; Luce, T. C.; Prater, R.

1998-11-01

Owing to relativistic mass shift in the cyclotron resonance condition, a simple and accurate interpolation formula for estimating the current drive efficiency, such as those(S.C. Chiu et al.), Nucl. Fusion 29, 2175 (1989).^,(D.A. Ehst and C.F.F. Karney, Nucl. Fusion 31), 1933 (1991). commonly used in FWCD, is not available in the case of ECCD. In this work, we model ECCD using the adjoint techniques. A semi-analytic adjoint function appropriate for general tokamak geometry is obtained using Fisch's relativistic collision model. Predictions of off-axis ECCD qualitatively and semi-quantitatively agrees with those of Cohen,(R.H. Cohen, Phys. Fluids 30), 2442 (1987). currently implemented in the raytracing code TORAY. The dependences of the current drive efficiency on the wave launch configuration and the plasma parameters will be presented. Strong absorption of the wave away from the resonance layer is shown to be an important factor in optimizing the off-axis ECCD for application to advanced tokamak operations.

Combined magnetic and kinetic control of advanced tokamak steady state scenarios based on semi-empirical modelling

DOE PAGES

Moreau, Didier; Artaud, J. F.; Ferron, John R.; ...

2015-05-01

This paper shows that semi-empirical data-driven models based on a twotime- scale approximation for the magnetic and kinetic control of advanced tokamak (AT) scenarios can be advantageously identified from simulated rather than real data, and used for control design. The method is applied to the combined control of the safety factor profile, q(x), and normalized pressure parameter, β N, using DIII-D parameters and actuators (on-axis co-current neutral beam injection (NBI) power, off axis co-current NBI power, electron cyclotron current drive power, and ohmic coil). The approximate plasma response model was identified from simulated data obtained using a rapidly converging plasmamore » transport code, METIS, which includes an MHD equilibrium and current diffusion solver, and combines plasma transport nonlinearity with 0-D scaling laws and 1.5-D ordinary differential equations. A number of open loop simulations were performed, in which the heating and current drive (H&CD) sources were randomly modulated around the typical values of a reference AT discharge on DIIID. Using these simulated data, a two-time-scale state space model was obtained for the coupled evolution of the poloidal flux profile and βN parameter, and a controller was synthesized based on the near-optimal ARTAEMIS algorithm [D. Moreau et al., Nucl. Fusion 53 (2013) 063020]. The paper discusses the results of closed-loop nonlinear simulations, using this controller for steady state AT operation. With feedforward plus feedback control, the steady state target q-profile and β N are satisfactorily tracked with a time scale of about ten seconds, despite large disturbances applied to the feedforward powers and plasma parameters. The effectiveness of the control algorithm is thus demonstrated for long pulse and steady state high-β N AT discharges. Its robustness with respect to disturbances of the H&CD actuators and of plasma parameters such as the H-factor, plasma density and effective charge, is also shown.« less

Advanced tokamak research with integrated modeling in JT-60 Upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayashi, N.

2010-05-15

Researches on advanced tokamak (AT) have progressed with integrated modeling in JT-60 Upgrade [N. Oyama et al., Nucl. Fusion 49, 104007 (2009)]. Based on JT-60U experimental analyses and first principle simulations, new models were developed and integrated into core, rotation, edge/pedestal, and scrape-off-layer (SOL)/divertor codes. The integrated models clarified complex and autonomous features in AT. An integrated core model was implemented to take account of an anomalous radial transport of alpha particles caused by Alfven eigenmodes. It showed the reduction in the fusion gain by the anomalous radial transport and further escape of alpha particles. Integrated rotation model showed mechanismsmore » of rotation driven by the magnetic-field-ripple loss of fast ions and the charge separation due to fast-ion drift. An inward pinch model of high-Z impurity due to the atomic process was developed and indicated that the pinch velocity increases with the toroidal rotation. Integrated edge/pedestal model clarified causes of collisionality dependence of energy loss due to the edge localized mode and the enhancement of energy loss by steepening a core pressure gradient just inside the pedestal top. An ideal magnetohydrodynamics stability code was developed to take account of toroidal rotation and clarified a destabilizing effect of rotation on the pedestal. Integrated SOL/divertor model clarified a mechanism of X-point multifaceted asymmetric radiation from edge. A model of the SOL flow driven by core particle orbits which partially enter the SOL was developed by introducing the ion-orbit-induced flow to fluid equations.« less

LIGKA: A linear gyrokinetic code for the description of background kinetic and fast particle effects on the MHD stability in tokamaks

NASA Astrophysics Data System (ADS)

Lauber, Ph.; Günter, S.; Könies, A.; Pinches, S. D.

2007-09-01

In a plasma with a population of super-thermal particles generated by heating or fusion processes, kinetic effects can lead to the additional destabilisation of MHD modes or even to additional energetic particle modes. In order to describe these modes, a new linear gyrokinetic MHD code has been developed and tested, LIGKA (linear gyrokinetic shear Alfvén physics) [Ph. Lauber, Linear gyrokinetic description of fast particle effects on the MHD stability in tokamaks, Ph.D. Thesis, TU München, 2003; Ph. Lauber, S. Günter, S.D. Pinches, Phys. Plasmas 12 (2005) 122501], based on a gyrokinetic model [H. Qin, Gyrokinetic theory and computational methods for electromagnetic perturbations in tokamaks, Ph.D. Thesis, Princeton University, 1998]. A finite Larmor radius expansion together with the construction of some fluid moments and specification to the shear Alfvén regime results in a self-consistent, electromagnetic, non-perturbative model, that allows not only for growing or damped eigenvalues but also for a change in mode-structure of the magnetic perturbation due to the energetic particles and background kinetic effects. Compared to previous implementations [H. Qin, mentioned above], this model is coded in a more general and comprehensive way. LIGKA uses a Fourier decomposition in the poloidal coordinate and a finite element discretisation in the radial direction. Both analytical and numerical equilibria can be treated. Integration over the unperturbed particle orbits is performed with the drift-kinetic HAGIS code [S.D. Pinches, Ph.D. Thesis, The University of Nottingham, 1996; S.D. Pinches et al., CPC 111 (1998) 131] which accurately describes the particles' trajectories. This allows finite-banana-width effects to be implemented in a rigorous way since the linear formulation of the model allows the exchange of the unperturbed orbit integration and the discretisation of the perturbed potentials in the radial direction. Successful benchmarks for toroidal Alfvén eigenmodes (TAEs) and kinetic Alfvén waves (KAWs) with analytical results, ideal MHD codes, drift-kinetic codes and other codes based on kinetic models are reported.

Numerical Solution of the Electron Heat Transport Equation and Physics-Constrained Modeling of the Thermal Conductivity via Sequential Quadratic Programming Optimization in Nuclear Fusion Plasmas

NASA Astrophysics Data System (ADS)

Paloma, Cynthia S.

The plasma electron temperature (Te) plays a critical role in a tokamak nu- clear fusion reactor since temperatures on the order of 108K are required to achieve fusion conditions. Many plasma properties in a tokamak nuclear fusion reactor are modeled by partial differential equations (PDE's) because they depend not only on time but also on space. In particular, the dynamics of the electron temperature is governed by a PDE referred to as the Electron Heat Transport Equation (EHTE). In this work, a numerical method is developed to solve the EHTE based on a custom finite-difference technique. The solution of the EHTE is compared to temperature profiles obtained by using TRANSP, a sophisticated plasma transport code, for specific discharges from the DIII-D tokamak, located at the DIII-D National Fusion Facility in San Diego, CA. The thermal conductivity (also called thermal diffusivity) of the electrons (Xe) is a plasma parameter that plays a critical role in the EHTE since it indicates how the electron temperature diffusion varies across the minor effective radius of the tokamak. TRANSP approximates Xe through a curve-fitting technique to match experimentally measured electron temperature profiles. While complex physics-based model have been proposed for Xe, there is a lack of a simple mathematical model for the thermal diffusivity that could be used for control design. In this work, a model for Xe is proposed based on a scaling law involving key plasma variables such as the electron temperature (Te), the electron density (ne), and the safety factor (q). An optimization algorithm is developed based on the Sequential Quadratic Programming (SQP) technique to optimize the scaling factors appearing in the proposed model so that the predicted electron temperature and magnetic flux profiles match predefined target profiles in the best possible way. A simulation study summarizing the outcomes of the optimization procedure is presented to illustrate the potential of the proposed modeling method.

SciDAC Center for Gyrokinetic Particle Simulation of Turbulent Transport in Burning Plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lin, Zhihong

2013-12-18

During the first year of the SciDAC gyrokinetic particle simulation (GPS) project, the GPS team (Zhihong Lin, Liu Chen, Yasutaro Nishimura, and Igor Holod) at the University of California, Irvine (UCI) studied the tokamak electron transport driven by electron temperature gradient (ETG) turbulence, and by trapped electron mode (TEM) turbulence and ion temperature gradient (ITG) turbulence with kinetic electron effects, extended our studies of ITG turbulence spreading to core-edge coupling. We have developed and optimized an elliptic solver using finite element method (FEM), which enables the implementation of advanced kinetic electron models (split-weight scheme and hybrid model) in the SciDACmore » GPS production code GTC. The GTC code has been ported and optimized on both scalar and vector parallel computer architectures, and is being transformed into objected-oriented style to facilitate collaborative code development. During this period, the UCI team members presented 11 invited talks at major national and international conferences, published 22 papers in peer-reviewed journals and 10 papers in conference proceedings. The UCI hosted the annual SciDAC Workshop on Plasma Turbulence sponsored by the GPS Center, 2005-2007. The workshop was attended by about fifties US and foreign researchers and financially sponsored several gradual students from MIT, Princeton University, Germany, Switzerland, and Finland. A new SciDAC postdoc, Igor Holod, has arrived at UCI to initiate global particle simulation of magnetohydrodynamics turbulence driven by energetic particle modes. The PI, Z. Lin, has been promoted to the Associate Professor with tenure at UCI.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ivanov, A. A., E-mail: aai@a5.kiam.ru; Martynov, A. A., E-mail: martynov@a5.kiam.ru; Medvedev, S. Yu., E-mail: medvedev@a5.kiam.ru

In the MHD tokamak plasma theory, the plasma pressure is usually assumed to be isotropic. However, plasma heating by neutral beam injection and RF heating can lead to a strong anisotropy of plasma parameters and rotation of the plasma. The development of MHD equilibrium theory taking into account the plasma inertia and anisotropic pressure began a long time ago, but until now it has not been consistently applied in computational codes for engineering calculations of the plasma equilibrium and evolution in tokamak. This paper contains a detailed derivation of the axisymmetric plasma equilibrium equation in the most general form (withmore » arbitrary rotation and anisotropic pressure) and description of the specialized version of the SPIDER code. The original method of calculation of the equilibrium with an anisotropic pressure and a prescribed rotational transform profile is proposed. Examples of calculations and discussion of the results are also presented.« less

Tokamak power reactor ignition and time dependent fractional power operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vold, E.L.; Mau, T.K.; Conn, R.W.

1986-06-01

A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transportmore » power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve.« less

Using AORSA to simulate helicon waves in DIII-D

NASA Astrophysics Data System (ADS)

Lau, C.; Jaeger, E. F.; Bertelli, N.; Berry, L. A.; Blazevski, D.; Green, D. L.; Murakami, M.; Park, J. M.; Pinsker, R. I.; Prater, R.

2015-12-01

Recent efforts have shown that helicon waves (fast waves at > 20ωci) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored, it will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects.

MHD and resonant instabilities in JT-60SA during current ramp-up with off-axis N-NB injection

NASA Astrophysics Data System (ADS)

Bierwage, A.; Toma, M.; Shinohara, K.

2017-12-01

The excitation of magnetohydrodynamic (MHD) and resonant instabilities and their effect on the plasma profiles during the current ramp-up phase of a beam-driven JT-60SA tokamak plasma is studied using the MHD-PIC hybrid code MEGA. In the simple scenario considered, the plasma is only driven by one negative-ion-based neutral beam, depositing 500 keV deuterons at 5 MW power off-axis at about mid-radius. The beam injection starts half-way in the ramp-up phase. Within 1 s, the beam-driven plasma current and fast ion pressure produce a configuration that is strongly unstable to rapidly growing MHD and resonant modes. Using MEGA, modes with low toroidal mode numbers in the range n = 1-4 are examined in detail and shown to cause substantial changes in the plasma profiles. The necessity to develop reduced models and incorporate the effects of such instabilities in integrated codes used to simulate the evolution of entire plasma discharges is discussed.

Profile control simulations and experiments on TCV: a controller test environment and results using a model-based predictive controller

NASA Astrophysics Data System (ADS)

Maljaars, E.; Felici, F.; Blanken, T. C.; Galperti, C.; Sauter, O.; de Baar, M. R.; Carpanese, F.; Goodman, T. P.; Kim, D.; Kim, S. H.; Kong, M.; Mavkov, B.; Merle, A.; Moret, J. M.; Nouailletas, R.; Scheffer, M.; Teplukhina, A. A.; Vu, N. M. T.; The EUROfusion MST1-team; The TCV-team

2017-12-01

The successful performance of a model predictive profile controller is demonstrated in simulations and experiments on the TCV tokamak, employing a profile controller test environment. Stable high-performance tokamak operation in hybrid and advanced plasma scenarios requires control over the safety factor profile (q-profile) and kinetic plasma parameters such as the plasma beta. This demands to establish reliable profile control routines in presently operational tokamaks. We present a model predictive profile controller that controls the q-profile and plasma beta using power requests to two clusters of gyrotrons and the plasma current request. The performance of the controller is analyzed in both simulation and TCV L-mode discharges where successful tracking of the estimated inverse q-profile as well as plasma beta is demonstrated under uncertain plasma conditions and the presence of disturbances. The controller exploits the knowledge of the time-varying actuator limits in the actuator input calculation itself such that fast transitions between targets are achieved without overshoot. A software environment is employed to prepare and test this and three other profile controllers in parallel in simulations and experiments on TCV. This set of tools includes the rapid plasma transport simulator RAPTOR and various algorithms to reconstruct the plasma equilibrium and plasma profiles by merging the available measurements with model-based predictions. In this work the estimated q-profile is merely based on RAPTOR model predictions due to the absence of internal current density measurements in TCV. These results encourage to further exploit model predictive profile control in experiments on TCV and other (future) tokamaks.

Improved two-point model for limiter scrape-off layer

NASA Astrophysics Data System (ADS)

Tokar, M. Z.; Kobayashi, M.; Feng, Y.

2004-10-01

An analytical model for a limiter scrape-off layer (SOL) is proposed, which takes self-consistently into account both conductive and convective contributions to the heat transport in SOL. The particle flows in the SOL main part are determined by considering the recycling of neutrals. The model allows us to interpret the results of numerical simulation by the code EMC3-EIRENE [Y. Feng, F. Sardei, P. Grigull, K. McCormick, J. Kisslinger, D. Reiter, and Y. Igitkhanov, Plasma Phys. Controlled Fusion 44, 611 (2002)] for the edge region of Tokamak Experiment for Technology Oriented Research (TEXTOR) [Proceedings of the 16th IEEE Symposium on Fusion Engineering, 1995 (Institute for Electrical and Electronics Engineers, Piscataway, NJ, 1995), p. 470].

Non-linear dynamics of compound sawteeth in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J.-H., E-mail: jae-heon.ahn@polytechnique.edu; Garbet, X.; Sabot, R.

2016-05-15

Compound sawteeth is studied with the XTOR-2F code. Non-linear full 3D magnetohydrodynamic simulations show that the plasma hot core is radially displaced and rotates during the partial crash, but is not fully expelled out of the q = 1 surface. Partial crashes occur when the radius of the q = 1 surface exceeds a critical value, at fixed poloidal beta. This critical value depends on the plasma elongation. The partial crash time is larger than the collapse time of an ordinary sawtooth, likely due to a weaker diamagnetic stabilization. This suggests that partial crashes result from a competition between destabilizing effects such as themore » q = 1 radius and diamagnetic stabilization.« less

Simultaneous Power Deposition Detection of Two EC Beams with the BIS Analysis in Moving TCV Plasmas

NASA Astrophysics Data System (ADS)

Curchod, L.; Pochelon, A.; Decker, J.; Felici, F.; Goodman, T. P.; Moret, J.-M.; Paley, J. I.

2009-11-01

Modulation of power amplitude is a widespread to determine the radial absorption profile of externally launched power in fusion plasmas. There are many techniques to analyze the plasma response to such a modulation. The break-in-slope (BIS) analysis can draw an estimated power deposition profile for each power step up. In this paper, the BIS analysis is used to monitor the power deposition location of one or two EC power beams simultaneously in a non-stationary plasma being displaced vertically in the TCV tokamak vessel. Except from radial discrepancies, the results have high time resolution and compare well with simulations from the R2D2-C3PO-LUKE ray-tracing and Fokker-Planck code suite.

Validation of the thermal transport model used for ITER startup scenario predictions with DIII-D experimental data

DOE PAGES

Casper, T. A.; Meyer, W. H.; Jackson, G. L.; ...

2010-12-08

We are exploring characteristics of ITER startup scenarios in similarity experiments conducted on the DIII-D Tokamak. In these experiments, we have validated scenarios for the ITER current ramp up to full current and developed methods to control the plasma parameters to achieve stability. Predictive simulations of ITER startup using 2D free-boundary equilibrium and 1D transport codes rely on accurate estimates of the electron and ion temperature profiles that determine the electrical conductivity and pressure profiles during the current rise. Here we present results of validation studies that apply the transport model used by the ITER team to DIII-D discharge evolutionmore » and comparisons with data from our similarity experiments.« less

Divertor electron temperature and impurity diffusion measurements with a spectrally resolved imaging radiometer.

PubMed

Clayton, D J; Jaworski, M A; Kumar, D; Stutman, D; Finkenthal, M; Tritz, K

2012-10-01

A divertor imaging radiometer (DIR) diagnostic is being studied to measure spatially and spectrally resolved radiated power P(rad)(λ) in the tokamak divertor. A dual transmission grating design, with extreme ultraviolet (~20-200 Å) and vacuum ultraviolet (~200-2000 Å) gratings placed side-by-side, can produce coarse spectral resolution over a broad wavelength range covering emission from impurities over a wide temperature range. The DIR can thus be used to evaluate the separate P(rad) contributions from different ion species and charge states. Additionally, synthetic spectra from divertor simulations can be fit to P(rad)(λ) measurements, providing a powerful code validation tool that can also be used to estimate electron divertor temperature and impurity transport.

A new hybrid-Lagrangian numerical scheme for gyrokinetic simulation of tokamak edge plasma

DOE PAGES

Ku, S.; Hager, R.; Chang, C. S.; ...

2016-04-01

In order to enable kinetic simulation of non-thermal edge plasmas at a reduced computational cost, a new hybrid-Lagrangian δf scheme has been developed that utilizes the phase space grid in addition to the usual marker particles, taking advantage of the computational strengths from both sides. The new scheme splits the particle distribution function of a kinetic equation into two parts. Marker particles contain the fast space-time varying, δf, part of the distribution function and the coarse-grained phase-space grid contains the slow space-time varying part. The coarse-grained phase-space grid reduces the memory-requirement and the computing cost, while the marker particles providemore » scalable computing ability for the fine-grained physics. Weights of the marker particles are determined by a direct weight evolution equation instead of the differential form weight evolution equations that the conventional delta-f schemes use. The particle weight can be slowly transferred to the phase space grid, thereby reducing the growth of the particle weights. The non-Lagrangian part of the kinetic equation – e.g., collision operation, ionization, charge exchange, heat-source, radiative cooling, and others – can be operated directly on the phase space grid. Deviation of the particle distribution function on the velocity grid from a Maxwellian distribution function – driven by ionization, charge exchange and wall loss – is allowed to be arbitrarily large. In conclusion, the numerical scheme is implemented in the gyrokinetic particle code XGC1, which specializes in simulating the tokamak edge plasma that crosses the magnetic separatrix and is in contact with the material wall.« less

A new hybrid-Lagrangian numerical scheme for gyrokinetic simulation of tokamak edge plasma

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ku, S.; Hager, R.; Chang, C. S.

In order to enable kinetic simulation of non-thermal edge plasmas at a reduced computational cost, a new hybrid-Lagrangian δf scheme has been developed that utilizes the phase space grid in addition to the usual marker particles, taking advantage of the computational strengths from both sides. The new scheme splits the particle distribution function of a kinetic equation into two parts. Marker particles contain the fast space-time varying, δf, part of the distribution function and the coarse-grained phase-space grid contains the slow space-time varying part. The coarse-grained phase-space grid reduces the memory-requirement and the computing cost, while the marker particles providemore » scalable computing ability for the fine-grained physics. Weights of the marker particles are determined by a direct weight evolution equation instead of the differential form weight evolution equations that the conventional delta-f schemes use. The particle weight can be slowly transferred to the phase space grid, thereby reducing the growth of the particle weights. The non-Lagrangian part of the kinetic equation – e.g., collision operation, ionization, charge exchange, heat-source, radiative cooling, and others – can be operated directly on the phase space grid. Deviation of the particle distribution function on the velocity grid from a Maxwellian distribution function – driven by ionization, charge exchange and wall loss – is allowed to be arbitrarily large. In conclusion, the numerical scheme is implemented in the gyrokinetic particle code XGC1, which specializes in simulating the tokamak edge plasma that crosses the magnetic separatrix and is in contact with the material wall.« less

A new hybrid-Lagrangian numerical scheme for gyrokinetic simulation of tokamak edge plasma

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ku, S., E-mail: sku@pppl.gov; Hager, R.; Chang, C.S.

In order to enable kinetic simulation of non-thermal edge plasmas at a reduced computational cost, a new hybrid-Lagrangian δf scheme has been developed that utilizes the phase space grid in addition to the usual marker particles, taking advantage of the computational strengths from both sides. The new scheme splits the particle distribution function of a kinetic equation into two parts. Marker particles contain the fast space-time varying, δf, part of the distribution function and the coarse-grained phase-space grid contains the slow space-time varying part. The coarse-grained phase-space grid reduces the memory-requirement and the computing cost, while the marker particles providemore » scalable computing ability for the fine-grained physics. Weights of the marker particles are determined by a direct weight evolution equation instead of the differential form weight evolution equations that the conventional delta-f schemes use. The particle weight can be slowly transferred to the phase space grid, thereby reducing the growth of the particle weights. The non-Lagrangian part of the kinetic equation – e.g., collision operation, ionization, charge exchange, heat-source, radiative cooling, and others – can be operated directly on the phase space grid. Deviation of the particle distribution function on the velocity grid from a Maxwellian distribution function – driven by ionization, charge exchange and wall loss – is allowed to be arbitrarily large. The numerical scheme is implemented in the gyrokinetic particle code XGC1, which specializes in simulating the tokamak edge plasma that crosses the magnetic separatrix and is in contact with the material wall.« less

EC assisted start-up experiments reproduction in FTU and AUG for simulations of the ITER case

DOE Office of Scientific and Technical Information (OSTI.GOV)

Granucci, G.; Ricci, D.; Farina, D.

The breakdown and plasma start-up in ITER are well known issues studied in the last few years in many tokamaks with the aid of calculation based on simplified modeling. The thickness of ITER metallic wall and the voltage limits of the Central Solenoid Power Supply strongly limit the maximum toroidal electric field achievable (0.3 V/m), well below the level used in the present generation of tokamaks. In order to have a safe and robust breakdown, the use of Electron Cyclotron Power to assist plasma formation and current rump up has been foreseen. This has raised attention on plasma formation phasemore » in presence of EC wave, especially in order to predict the required power for a robust breakdown in ITER. Few detailed theory studies have been performed up to nowadays, due to the complexity of the problems. A simplified approach, extended from that proposed in ref[1] has been developed including a impurity multispecies distribution and an EC wave propagation and absorption based on GRAY code. This integrated model (BK0D) has been benchmarked on ohmic and EC assisted experiments on FTU and AUG, finding the key aspects for a good reproduction of data. On the basis of this, the simulation has been devoted to understand the best configuration for ITER case. The dependency of impurity distribution content and neutral gas pressure limits has been considered. As results of the analysis a reasonable amount of power (1 - 2 MW) seems to be enough to extend in a significant way the breakdown and current start up capability of ITER. The work reports the FTU data reproduction and the ITER case simulations.« less

Total fluid pressure imbalance in the scrape-off layer of tokamak plasmas

DOE PAGES

Churchill, Randy M.; Canik, John M.; Chang, C. S.; ...

2017-03-10

Simulations using the fully kinetic neoclassical code XGCa (X-point included guiding-center axisymmetric) were undertaken to explore the impact of kinetic effects on scrape-off layer (SOL) physics in DIII-D H-mode plasmas. XGCa is a total-f, gyrokinetic code which self-consistently calculates the axisymmetric electrostatic potential and plasma dynamics, and includes modules for Monte Carlo neutral transport. Previously presented XGCa results showed several noteworthy features, including large variations of ion density and pressure along field lines in the SOL, experimentally relevant levels of SOL parallel ion flow (Mach number similar to 0.5), skewed ion distributions near the sheath entrance leading to subsonic flowmore » there, and elevated sheath potentials (Churchill 2016 Nucl. Mater. Energy 1-6). In this paper, we explore in detail the question of pressure balance in the SOL, as it was observed in the simulation that there was a large deviation from a simple total pressure balance (the sum of ion and electron static pressure plus ion inertia). It will be shown that both the contributions from the ion viscosity (driven by ion temperature anisotropy) and neutral source terms can be substantial, and should be retained in the parallel momentum equation in the SOL, but still falls short of accounting for the observed fluid pressure imbalance in the XGCa simulation results.« less

Material Surface Damage under High Pulse Loads Typical for ELM Bursts and Disruptions in ITER

NASA Astrophysics Data System (ADS)

Landman, I. S.; Pestchanyi, S. E.; Safronov, V. M.; Bazylev, B. N.; Garkusha, I. E.

The divertor armour material for the tokamak ITER will probably be carbon manufactured as fibre composites (CFC) and tungsten as either brush-like structures or thin plates. Disruptive pulse loads where the heat deposition Q may reach 102 MJ/m 2 on a time scale Ïä of 3 ms, or operation in the ELMy H-mode at repetitive loads with Q âe 1/4 3 MJ/m2 and Ïä âe 1/4 0.3 ms, deteriorate armour performance. This work surveys recent numerical and experimental investigations of erosion mechanisms at these off-normal regimes carried out at FZK, TRINITI, and IPP-Kharkov. The modelling uses the anisotropic thermomechanics code PEGASUS-3D for the simulation of CFC brittle destruction, the surface melt motion code MEMOS-1.5D for tungsten targets, and the radiation-magnetohydrodynamics code FOREV-2D for calculating the plasma impact and simulating the heat loads for the ITER regime. Experiments aimed at validating these codes are being carried out at the plasma gun facilities MK-200UG, QSPA-T, and QSPA-Kh50 which produce powerful streams of hydrogen plasma with Q = 10–30 MJ/m2 and Ïä = 0.03–0.5 ms. Essential results are, for CFC targets, the experiments at high heat loads and the development of a local overheating model incorporated in PEGASUS-3D, and for the tungsten targets the analysis of evaporation- and melt motion erosion on the base of MEMOS-1.5D calculations for repetitive ELMs.

Edge-localized-modes in tokamaks

DOE PAGES

Leonard, Anthony W.

2014-09-11

Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heatmore » flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. As a result, encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.« less

Neoclassical toroidal viscosity calculations in tokamaks using a δf Monte Carlo simulation and their verifications.

PubMed

Satake, S; Park, J-K; Sugama, H; Kanno, R

2011-07-29

Neoclassical toroidal viscosities (NTVs) in tokamaks are investigated using a δf Monte Carlo simulation, and are successfully verified with a combined analytic theory over a wide range of collisionality. A Monte Carlo simulation has been required in the study of NTV since the complexities in guiding-center orbits of particles and their collisions cannot be fully investigated by any means of analytic theories alone. Results yielded the details of the complex NTV dependency on particle precessions and collisions, which were predicted roughly in a combined analytic theory. Both numerical and analytic methods can be utilized and extended based on these successful verifications.

Core turbulence behavior moving from ion-temperature-gradient regime towards trapped-electron-mode regime in the ASDEX Upgrade tokamak and comparison with gyrokinetic simulation

NASA Astrophysics Data System (ADS)

Happel, T.; Navarro, A. Bañón; Conway, G. D.; Angioni, C.; Bernert, M.; Dunne, M.; Fable, E.; Geiger, B.; Görler, T.; Jenko, F.; McDermott, R. M.; Ryter, F.; Stroth, U.

2015-03-01

Additional electron cyclotron resonance heating (ECRH) is used in an ion-temperature-gradient instability dominated regime to increase R / L Te in order to approach the trapped-electron-mode instability regime. The radial ECRH deposition location determines to a large degree the effect on R / L Te . Accompanying scale-selective turbulence measurements at perpendicular wavenumbers between k⊥ = 4-18 cm-1 (k⊥ρs = 0.7-4.2) show a pronounced increase of large-scale density fluctuations close to the ECRH radial deposition location at mid-radius, along with a reduction in phase velocity of large-scale density fluctuations. Measurements are compared with results from linear and non-linear flux-matched gyrokinetic (GK) simulations with the gyrokinetic code GENE. Linear GK simulations show a reduction of phase velocity, indicating a pronounced change in the character of the dominant instability. Comparing measurement and non-linear GK simulation, as a central result, agreement is obtained in the shape of radial turbulence level profiles. However, the turbulence intensity is increasing with additional heating in the experiment, while gyrokinetic simulations show a decrease.

FLiT: a field line trace code for magnetic confinement devices

NASA Astrophysics Data System (ADS)

Innocente, P.; Lorenzini, R.; Terranova, D.; Zanca, P.

2017-04-01

This paper presents a field line tracing code (FLiT) developed to study particle and energy transport as well as other phenomena related to magnetic topology in reversed-field pinch (RFP) and tokamak experiments. The code computes magnetic field lines in toroidal geometry using curvilinear coordinates (r, ϑ, ϕ) and calculates the intersections of these field lines with specified planes. The code also computes the magnetic and thermal diffusivity due to stochastic magnetic field in the collisionless limit. Compared to Hamiltonian codes, there are no constraints on the magnetic field functional formulation, which allows the integration of whichever magnetic field is required. The code uses the magnetic field computed by solving the zeroth-order axisymmetric equilibrium and the Newcomb equation for the first-order helical perturbation matching the edge magnetic field measurements in toroidal geometry. Two algorithms are developed to integrate the field lines: one is a dedicated implementation of a first-order semi-implicit volume-preserving integration method, and the other is based on the Adams-Moulton predictor-corrector method. As expected, the volume-preserving algorithm is accurate in conserving divergence, but slow because the low integration order requires small amplitude steps. The second algorithm proves to be quite fast and it is able to integrate the field lines in many partially and fully stochastic configurations accurately. The code has already been used to study the core and edge magnetic topology of the RFX-mod device in both the reversed-field pinch and tokamak magnetic configurations.

140 GHz EC waves propagation and absorption for normal/oblique injection on FTU tokamak

NASA Astrophysics Data System (ADS)

Nowak, S.; Airoldi, A.; Bruschi, A.; Buratti, P.; Cirant, S.; Gandini, F.; Granucci, G.; Lazzaro, E.; Panaccione, L.; Ramponi, G.; Simonetto, A.; Sozzi, C.; Tudisco, O.; Zerbini, M.

1999-09-01

Most of the interest in ECRH experiments is linked to the high localization of EC waves absorption in well known portions of the plasma volume. In order to take full advantage of this capability a reliable code has been developed for beam tracing and absorption calculations. The code is particularly important for oblique (poloidal and toroidal) injection, when the absorbing layer is not simply dependent on the position of the EC resonance only. An experimental estimate of the local heating power density is given by the jump in the time derivative of the local electron pressure at the switching ON of the gyrotron power. The evolution of the temperature profile increase (from ECE polychromator) during the nearly adiabatic phase is also considered for ECRH profile reconstruction. An indirect estimate of optical thickness and of the overall absorption coefficient is given by the measure of the residual e.m. power at the tokamak walls. Beam tracing code predictions of the power deposition profile are compared with experimental estimates. The impact of the finite spatial resolution of the temperature diagnostic on profile reconstruction is also discussed.

Confinement properties of tokamak plasmas with extended regions of low magnetic shear

NASA Astrophysics Data System (ADS)

Graves, J. P.; Cooper, W. A.; Kleiner, A.; Raghunathan, M.; Neto, E.; Nicolas, T.; Lanthaler, S.; Patten, H.; Pfefferle, D.; Brunetti, D.; Lutjens, H.

2017-10-01

Extended regions of low magnetic shear can be advantageous to tokamak plasmas. But the core and edge can be susceptible to non-resonant ideal fluctuations due to the weakened restoring force associated with magnetic field line bending. This contribution shows how saturated non-linear phenomenology, such as 1 / 1 Long Lived Modes, and Edge Harmonic Oscillations associated with QH-modes, can be modelled accurately using the non-linear stability code XTOR, the free boundary 3D equilibrium code VMEC, and non-linear analytic theory. That the equilibrium approach is valid is particularly valuable because it enables advanced particle confinement studies to be undertaken in the ordinarily difficult environment of strongly 3D magnetic fields. The VENUS-LEVIS code exploits the Fourier description of the VMEC equilibrium fields, such that full Lorenzian and guiding centre approximated differential operators in curvilinear angular coordinates can be evaluated analytically. Consequently, the confinement properties of minority ions such as energetic particles and high Z impurities can be calculated accurately over slowing down timescales in experimentally relevant 3D plasmas.

Shutdown Dose Rate Analysis for the long-pulse D-D Operation Phase in KSTAR

NASA Astrophysics Data System (ADS)

Park, Jin Hun; Han, Jung-Hoon; Kim, D. H.; Joo, K. S.; Hwang, Y. S.

2017-09-01

KSTAR is a medium size fully superconducting tokamak. The deuterium-deuterium (D-D) reaction in the KSTAR tokamak generates neutrons with a peak yield of 3.5x1016 per second through a pulse operation of 100 seconds. The effect of neutron generation from full D-D high power KSTAR operation mode to the machine, such as activation, shutdown dose rate, and nuclear heating, are estimated for an assurance of safety during operation, maintenance, and machine upgrade. The nuclear heating of the in-vessel components, and neutron activation of the surrounding materials have been investigated. The dose rates during operation and after shutdown of KSTAR have been calculated by a 3D CAD model of KSTAR with the Monte Carlo code MCNP5 (neutron flux and decay photon), the inventory code FISPACT (activation and decay photon) and the FENDL 2.1 nuclear data library.

Burn Control Mechanisms in Tokamaks

NASA Astrophysics Data System (ADS)

Hill, M. A.; Stacey, W. M.

2015-11-01

Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

Development of FullWave : Hot Plasma RF Simulation Tool

NASA Astrophysics Data System (ADS)

Svidzinski, Vladimir; Kim, Jin-Soo; Spencer, J. Andrew; Zhao, Liangji; Galkin, Sergei

2017-10-01

Full wave simulation tool, modeling RF fields in hot inhomogeneous magnetized plasma, is being developed. The wave equations with linearized hot plasma dielectric response are solved in configuration space on adaptive cloud of computational points. The nonlocal hot plasma dielectric response is formulated in configuration space without limiting approximations by calculating the plasma conductivity kernel based on the solution of the linearized Vlasov equation in inhomogeneous magnetic field. This approach allows for better resolution of plasma resonances, antenna structures and complex boundaries. The formulation of FullWave and preliminary results will be presented: construction of the finite differences for approximation of derivatives on adaptive cloud of computational points; model and results of nonlocal conductivity kernel calculation in tokamak geometry; results of 2-D full wave simulations in the cold plasma model in tokamak geometry using the formulated approach; results of self-consistent calculations of hot plasma dielectric response and RF fields in 1-D mirror magnetic field; preliminary results of self-consistent simulations of 2-D RF fields in tokamak using the calculated hot plasma conductivity kernel; development of iterative solver for wave equations. Work is supported by the U.S. DOE SBIR program.

Gyrofluid Modeling of Turbulent, Kinetic Physics

NASA Astrophysics Data System (ADS)

Despain, Kate Marie

2011-12-01

Gyrofluid models to describe plasma turbulence combine the advantages of fluid models, such as lower dimensionality and well-developed intuition, with those of gyrokinetics models, such as finite Larmor radius (FLR) effects. This allows gyrofluid models to be more tractable computationally while still capturing much of the physics related to the FLR of the particles. We present a gyrofluid model derived to capture the behavior of slow solar wind turbulence and describe the computer code developed to implement the model. In addition, we describe the modifications we made to a gyrofluid model and code that simulate plasma turbulence in tokamak geometries. Specifically, we describe a nonlinear phase mixing phenomenon, part of the E x B term, that was previously missing from the model. An inherently FLR effect, it plays an important role in predicting turbulent heat flux and diffusivity levels for the plasma. We demonstrate this importance by comparing results from the updated code to studies done previously by gyrofluid and gyrokinetic codes. We further explain what would be necessary to couple the updated gyrofluid code, gryffin, to a turbulent transport code, thus allowing gryffin to play a role in predicting profiles for fusion devices such as ITER and to explore novel fusion configurations. Such a coupling would require the use of Graphical Processing Units (GPUs) to make the modeling process fast enough to be viable. Consequently, we also describe our experience with GPU computing and demonstrate that we are poised to complete a gryffin port to this innovative architecture.

Advanced Divertor Design and Application under Modern Superconducting Tokamak Constraints

NASA Astrophysics Data System (ADS)

Covele, Brent; Kotschenreuther, Mike; Mahajan, Swadesh; Valanju, Prashant

2013-10-01

With current ITER projections already predicting divertor exhaust heat loads in the 5-10 MW/m2 range, i.e. at the maximum tolerance, it is clear that the divertor heat load problem will only be exacerbated for future superconducting tokamaks, as well as perhaps some modern tokamaks today. Thus, an advanced divertor, such as the X-Divertor (XD), Super-X Divertor (SXD), or Snowflake (SF) will become a virtual necessity to reduce incident heat flux at the target plates. Using the 2D magnetic equilibrium code CORSICA, we explore the possibilities of creating an advanced divertor for a next-generation superconducting tokamak (Ip = 15 MA, BT = 5.3 T, R = 6.2 m) under nominal engineering constraints. Advanced divertors were achieved with no in-vessel PF coils, PF current densities below 30 MA/m2, and vertical maintenance access, all of which are favorable conditions for tokamaks today. Both the XD and SF divertors are readily achievable while maintaining core plasma performance, and the advantages and disadvantages of each are discussed in turn. Some thought is given as to how the divertor cassette will need to be modified to accommodate advanced divertors. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

Simulation of density fluctuations before the L-H transition for Hydrogen and Deuterium plasmas in the DIII-D tokamak using the BOUT++ code

NASA Astrophysics Data System (ADS)

Wang, Y. M.; Xu, X. Q.; Yan, Z.; Mckee, G. R.; Grierson, B. A.; Xia, T. Y.; Gao, X.

2018-02-01

A six-field two-fluid model has been used to simulate density fluctuations. The equilibrium is generated by experimental measurements for both Deuterium (D) and Hydrogen (H) plasmas at the lowest densities of DIII-D low to high confinement (L-H) transition experiments. In linear simulations, the unstable modes are found to be resistive ballooning modes with the most unstable mode number n  =  30 or k_θρ_i˜0.12 . The ion diamagnetic drift and E× B convection flow are balanced when the radial electric field (E r ) is calculated from the pressure profile without net flow. The curvature drift plays an important role in this stage. Two poloidally counter propagating modes are found in the nonlinear simulation of the D plasma at electron density n_e˜1.5×1019 m-3 near the separatrix while a single ion mode is found in the H plasma at the similar lower density, which are consistent with the experimental results measured by the beam emission spectroscopy (BES) diagnostic on the DIII-D tokamak. The frequency of the electron modes and the ion modes are about 40 kHz and 10 kHz respectively. The poloidal wave number k_θ is about 0.2 cm -1 (k_θρ_i˜0.05 ) for both ion and electron modes. The particle flux, ion and electron heat fluxes are  ˜3.5-6 times larger for the H plasma than the D plasma, which makes it harder to achieve H-mode for the same heating power. The change of the atomic mass number A from 2 to 1 using D plasma equilibrium make little difference on the flux. Increase the electric field will suppress the density fluctuation. The electric field scan and ion mass scan results show that the dual-mode results primarily from differences in the profiles rather than the ion mass.

Validating the energy transport modeling of the DIII-D and EAST ramp up experiments using TSC

NASA Astrophysics Data System (ADS)

Liu, Li; Guo, Yong; Chan, Vincent; Mao, Shifeng; Wang, Yifeng; Pan, Chengkang; Luo, Zhengping; Zhao, Hailin; Ye, Minyou

2017-06-01

The confidence in ramp up scenario design of the China fusion engineering test reactor (CFETR) can be significantly enhanced using validated transport models to predict the current profile and temperature profile. In the tokamak simulation code (TSC), two semi-empirical energy transport models (the Coppi-Tang (CT) and BGB model) and three theory-based models (the GLF23, MMM95 and CDBM model) are investigated on the CFETR relevant ramp up discharges, including three DIII-D ITER-like ramp up discharges and one EAST ohmic discharge. For the DIII-D discharges, all the transport models yield dynamic {{\\ell}\\text{i}} within +/- 0.15 deviations except for some time points where the experimental fluctuation is very strong. All the models agree with the experimental {β\\text{p}} except that the CT model strongly overestimates {β\\text{p}} in the first half of ramp up phase. When applying the CT, CDBM and GLF23 model to estimate the internal flux, they show maximum deviations of more than 10% because of inaccuracies in the temperature profile predictions, while the BGB model performs best on the internal flux. Although all the models fall short in reproducing the dynamic {{\\ell}\\text{i}} evolution for the EAST tokamak, the result of the BGB model is the closest to the experimental {{\\ell}\\text{i}} . Based on these comparisons, we conclude that the BGB model is the most consistent among these models for simulating CFETR ohmic ramp-up. The CT model with improvement for better simulation of the temperature profiles in the first half of ramp up phase will also be attractive. For the MMM95, GLF23 and CDBM model, better prediction of the edge temperature will improve the confidence for CFETR L-mode simulation. Conclusive validation of any transport model will require extensive future investigation covering a larger variety discharges.

Identifying microturbulence regimes in a TCV discharge making use of physical constraints on particle and heat fluxes

DOE PAGES

Mariani, Alberto; Brunner, S.; Dominski, J.; ...

2018-01-17

Reducing the uncertainty on physical input parameters derived from experimental measurements is essential towards improving the reliability of gyrokinetic turbulence simulations. This can be achieved by introducing physical constraints. Amongst them, the zero particle flux condition is considered here. A first attempt is also made to match as well the experimental ion/electron heat flux ratio. This procedure is applied to the analysis of a particular Tokamak à Configuration Variable discharge. A detailed reconstruction of the zero particle flux hyper-surface in the multi-dimensional physical parameter space at fixed time of the discharge is presented, including the effect of carbon as themore » main impurity. Both collisionless and collisional regimes are considered. Hyper-surface points within the experimental error bars are found. In conclusion, the analysis is done performing gyrokinetic simulations with the local version of the GENE code, computing the fluxes with a Quasi-Linear (QL) model and validating the QL results with non-linear simulations in a subset of cases.« less

Testing the Porcelli Sawtooth Trigger Module

NASA Astrophysics Data System (ADS)

Bateman, G.; Nave, M. F. F.; Parail, V.

2005-10-01

The Porcelli sawtooth trigger model [1] is implemented as a module for the National Transport Code Collaboration Module Library [2] and is tested using BALDUR and JETTO integrated modeling simulations of JET and other tokamak discharges. Statistical techniques are used to compute the average sawtooth period and the random scatter in sawtooth periods obtained during selected time intervals in the simulations compared with the corresponding statistical measures obtained from experimental data. It is found that the results are affected systematically by the fraction of magnetic reconnection during each sawtooth crash and by the model that is used for transport within the sawtooth mixing region. The physical processes that affect the sawtooth cycle in the simulations are found to involve an interaction among magnetic diffusion, reheating within the sawtooth mixing region, the instabilities that trigger a sawtooth crash in the Porcelli model, and the magnetic reconnection produced by each sawtooth crash. [1] F. Porcelli, et al., Plasma Phys. Contol. Fusion 38 (1996) 2163. [2] A.H. Kritz, et al., Comput. Phys. Commun. 164 (2004) 108; http://w3.pppl.gov/NTCC. Supported by DOE DE-FG02-92-ER-54141.

Identifying microturbulence regimes in a TCV discharge making use of physical constraints on particle and heat fluxes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mariani, Alberto; Brunner, S.; Dominski, J.

Reducing the uncertainty on physical input parameters derived from experimental measurements is essential towards improving the reliability of gyrokinetic turbulence simulations. This can be achieved by introducing physical constraints. Amongst them, the zero particle flux condition is considered here. A first attempt is also made to match as well the experimental ion/electron heat flux ratio. This procedure is applied to the analysis of a particular Tokamak à Configuration Variable discharge. A detailed reconstruction of the zero particle flux hyper-surface in the multi-dimensional physical parameter space at fixed time of the discharge is presented, including the effect of carbon as themore » main impurity. Both collisionless and collisional regimes are considered. Hyper-surface points within the experimental error bars are found. In conclusion, the analysis is done performing gyrokinetic simulations with the local version of the GENE code, computing the fluxes with a Quasi-Linear (QL) model and validating the QL results with non-linear simulations in a subset of cases.« less

Theory-based transport simulations of TFTR L-mode temperature profiles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bateman, G.

1992-03-01

The temperature profiles from a selection of Tokamak Fusion Test Reactor (TFTR) L-mode discharges (17{ital th} {ital European} {ital Conference} {ital on} {ital Controlled} {ital Fusion} {ital and} {ital Plasma} {ital Heating}, Amsterdam, 1990 (EPS, Petit-Lancy, Switzerland, 1990, p. 114)) are simulated with the 1 (1)/(2) -D baldur transport code (Comput. Phys. Commun. {bold 49}, 275 (1988)) using a combination of theoretically derived transport models, called the Multi-Mode Model (Comments Plasma Phys. Controlled Fusion {bold 11}, 165 (1988)). The present version of the Multi-Mode Model consists of effective thermal diffusivities resulting from trapped electron modes and ion temperature gradient ({eta}{submore » {ital i}}) modes, which dominate in the core of the plasma, together with resistive ballooning modes, which dominate in the periphery. Within the context of this transport model and the TFTR simulations reported here, the scaling of confinement with heating power comes from the temperature dependence of the {eta}{sub {ital i}} and trapped electron modes, while the scaling with current comes mostly from resistive ballooning modes.« less

Evidence of a New Instability in Gyrokinetic Simulations of LAPD Plasmas

NASA Astrophysics Data System (ADS)

Terry, P. W.; Pueschel, M. J.; Rossi, G.; Jenko, F.; Told, D.; Carter, T. A.

2015-11-01

Recent experiments at the LArge Plasma Device (LAPD) have focused on structure formation driven by density and temperature gradients. A central difference relative to typical, tokamak-like plasmas stems from the linear geometry and absence of background magnetic shear. At sufficiently high β, strong excitation of parallel (compressional) magnetic fluctuations was observed. Here, linear and nonlinear simulations with the Gene code are used to demonstrate that these findings can be explained through the linear excitation of a Gradient-driven Drift Coupling mode (GDC). This recently-discovered instability, unlike other drift waves, relies on the grad-B drift due to parallel magnetic fluctuations in lieu of a parallel electron response, and can be driven by density or temperature gradients. The linear properties of the GDC for LAPD parameters are studied in detail, and the corresponding turbulence is investigated. It is found that, despite the very large collisionality in the experiment, many properties are recovered fairly well in the simulations. In addition to confirming the existence of the GDC, this opens up interesting questions regarding GDC activity in astrophysical and space plasmas. Supported by USDOE.

What happens to full-f gyrokinetic transport and turbulence in a toroidal wedge simulation?

DOE PAGES

Kim, Kyuho; Chang, C. S.; Seo, Janghoon; ...

2017-01-24

Here, in order to save the computing time or to fit the simulation size into a limited computing hardware in a gyrokinetic turbulence simulation of a tokamak plasma, a toroidal wedge simulation may be utilized in which only a partial toroidal section is modeled with a periodic boundary condition in the toroidal direction. The most severe restriction in the wedge simulation is expected to be in the longest wavelength turbulence, i.e., ion temperature gradient (ITG) driven turbulence. The global full-f gyrokinetic code XGC1 is used to compare the transport and turbulence properties from a toroidal wedge simulation against the fullmore » torus simulation in an ITG unstable plasma in a model toroidal geometry. It is found that (1) the convergence study in the wedge number needs to be conducted all the way down to the full torus in order to avoid a false convergence, (2) a reasonably accurate simulation can be performed if the correct wedge number N can be identified, (3) the validity of a wedge simulation may be checked by performing a wave-number spectral analysis of the turbulence amplitude |δΦ| and assuring that the variation of δΦ between the discrete kθ values is less than 25% compared to the peak |δΦ|, and (4) a frequency spectrum may not be used for the validity check of a wedge simulation.« less

What happens to full-f gyrokinetic transport and turbulence in a toroidal wedge simulation?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kyuho; Chang, C. S.; Seo, Janghoon

Here, in order to save the computing time or to fit the simulation size into a limited computing hardware in a gyrokinetic turbulence simulation of a tokamak plasma, a toroidal wedge simulation may be utilized in which only a partial toroidal section is modeled with a periodic boundary condition in the toroidal direction. The most severe restriction in the wedge simulation is expected to be in the longest wavelength turbulence, i.e., ion temperature gradient (ITG) driven turbulence. The global full-f gyrokinetic code XGC1 is used to compare the transport and turbulence properties from a toroidal wedge simulation against the fullmore » torus simulation in an ITG unstable plasma in a model toroidal geometry. It is found that (1) the convergence study in the wedge number needs to be conducted all the way down to the full torus in order to avoid a false convergence, (2) a reasonably accurate simulation can be performed if the correct wedge number N can be identified, (3) the validity of a wedge simulation may be checked by performing a wave-number spectral analysis of the turbulence amplitude |δΦ| and assuring that the variation of δΦ between the discrete kθ values is less than 25% compared to the peak |δΦ|, and (4) a frequency spectrum may not be used for the validity check of a wedge simulation.« less

Wall touching kink mode calculations with the M3D code

NASA Astrophysics Data System (ADS)

Breslau, J. A.

2014-10-01

In recent years there have been a number of results published concerning the transient vessel currents and forces occurring during a tokamak VDE, as predicted by simulations with the nonlinear MHD code M3D. The nature of the simulations is such that these currents and forces occur at the boundary of the computational domain, making the proper choice of boundary conditions critical to the reliability of the results. The M3D boundary condition includes the prescription that the normal component of the velocity vanish at the wall. It has been argued that this prescription invalidates the calculations because it would seem to rule out the possibility of advection of plasma surface currents into the wall. This claim has been tested by applying M3D to an idealized case - a kink-unstable plasma column - in order to abstract the essential physics from the complications involved in the attempt to model real devices. While comparison of the results is complicated by effects arising from the higher dimensionality and complexity of M3D, we have verified that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the ``Hiro'' currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.

Nonlinear simulations with and computational issues for NIMROD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sovinec, C R

The NIMROD (Non-Ideal Magnetohydrodynamics with Rotation, Open Discussion) code development project was commissioned by the US Department of Energy in February, 1996 to provide the fusion research community with a computational tool for studying low-frequency behavior in experiments. Specific problems of interest include the neoclassical evolution of magnetic islands and the nonlinear behavior of tearing modes in the presence of rotation and nonideal walls in tokamaks; they also include topics relevant to innovative confinement concepts such as magnetic turbulence. Besides having physics models appropriate for these phenomena, an additional requirement is the ability to perform the computations in realistic geometries.more » The NIMROD Team is using contemporary management and computational methods to develop a computational tool for investigating low-frequency behavior in plasma fusion experiments. The authors intend to make the code freely available, and are taking steps to make it as easy to learn and use as possible. An example application for NIMROD is the nonlinear toroidal RFP simulation--the first in a series to investigate how toroidal geometry affects MHD activity in RFPs. Finally, the most important issue facing the project is execution time, and they are exploring better matrix solvers and a better parallel decomposition to address this.« less

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

NASA Astrophysics Data System (ADS)

Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

2015-09-01

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chrystal, C.; Grierson, B. A.; Staebler, G. M.

Here, experiments at the DIII-D tokamak have used dimensionless parameter scans to investigate the dependencies of intrinsic torque and momentum transport in order to inform a prediction of the rotation profile in ITER. Measurements of intrinsic torque profiles and momentum confinement time in dimensionless parameter scans of normalized gyroradius and collisionality are used to predict the amount of intrinsic rotation in the pedestal of ITER. Additional scans of T e/T i and safety factor are used to determine the accuracy of momentum flux predictions of the quasi-linear gyrokinetic code TGLF. In these scans, applications of modulated torque are used tomore » measure the incremental momentum diffusivity, and results are consistent with the E x B shear suppression of turbulent transport. These incremental transport measurements are also compared with the TGLF results. In order to form a prediction of the rotation profile for ITER, the pedestal prediction is used as a boundary condition to a simulation that uses TGLF to determine the transport in the core of the plasma. The predicted rotation is ≈20 krad/s in the core, lower than in many current tokamak operating scenarios. TGLF predictions show that this rotation is still significant enough to have a strong effect on confinement via E x B shear.« less

Predicting rotation for ITER via studies of intrinsic torque and momentum transport in DIII-D

DOE PAGES

Chrystal, C.; Grierson, B. A.; Staebler, G. M.; ...

2017-03-30

Here, experiments at the DIII-D tokamak have used dimensionless parameter scans to investigate the dependencies of intrinsic torque and momentum transport in order to inform a prediction of the rotation profile in ITER. Measurements of intrinsic torque profiles and momentum confinement time in dimensionless parameter scans of normalized gyroradius and collisionality are used to predict the amount of intrinsic rotation in the pedestal of ITER. Additional scans of T e/T i and safety factor are used to determine the accuracy of momentum flux predictions of the quasi-linear gyrokinetic code TGLF. In these scans, applications of modulated torque are used tomore » measure the incremental momentum diffusivity, and results are consistent with the E x B shear suppression of turbulent transport. These incremental transport measurements are also compared with the TGLF results. In order to form a prediction of the rotation profile for ITER, the pedestal prediction is used as a boundary condition to a simulation that uses TGLF to determine the transport in the core of the plasma. The predicted rotation is ≈20 krad/s in the core, lower than in many current tokamak operating scenarios. TGLF predictions show that this rotation is still significant enough to have a strong effect on confinement via E x B shear.« less

Bootstrap Current for the Edge Pedestal Plasma in a Diverted Tokamak Geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koh, S.; Chang, C. S.; Ku, S.

The edge bootstrap current plays a critical role in the equilibrium and stability of the steep edge pedestal plasma. The pedestal plasma has an unconventional and difficult neoclassical property, as compared with the core plasma. It has a narrow passing particle region in velocity space that can be easily modified or destroyed by Coulomb collisions. At the same time, the edge pedestal plasma has steep pressure and electrostatic potential gradients whose scale-lengths are comparable with the ion banana width, and includes a magnetic separatrix surface, across which the topological properties of the magnetic field and particle orbits change abruptly. Amore » driftkinetic particle code XGC0, equipped with a mass-momentum-energy conserving collision operator, is used to study the edge bootstrap current in a realistic diverted magnetic field geometry with a self-consistent radial electric field. When the edge electrons are in the weakly collisional banana regime, surprisingly, the present kinetic simulation confirms that the existing analytic expressions [represented by O. Sauter et al. , Phys. Plasmas 6 , 2834 (1999)] are still valid in this unconventional region, except in a thin radial layer in contact with the magnetic separatrix. The agreement arises from the dominance of the electron contribution to the bootstrap current compared with ion contribution and from a reasonable separation of the trapped-passing dynamics without a strong collisional mixing. However, when the pedestal electrons are in plateau-collisional regime, there is significant deviation of numerical results from the existing analytic formulas, mainly due to large effective collisionality of the passing and the boundary layer trapped particles in edge region. In a conventional aspect ratio tokamak, the edge bootstrap current from kinetic simulation can be significantly less than that from the Sauter formula if the electron collisionality is high. On the other hand, when the aspect ratio is close to unity, the collisional edge bootstrap current can be significantly greater than that from the Sauter formula. Rapid toroidal rotation of the magnetic field lines at the high field side of a tight aspect-ratio tokamak is believed to be the cause of the different behavior. A new analytic fitting formula, as a simple modification to the Sauter formula, is obtained to bring the analytic expression to a better agreement with the edge kinetic simulation results« less

Bootstrap current for the edge pedestal plasma in a diverted tokamak geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koh, S.; Choe, W.; Chang, C. S.

The edge bootstrap current plays a critical role in the equilibrium and stability of the steep edge pedestal plasma. The pedestal plasma has an unconventional and difficult neoclassical property, as compared with the core plasma. It has a narrow passing particle region in velocity space that can be easily modified or destroyed by Coulomb collisions. At the same time, the edge pedestal plasma has steep pressure and electrostatic potential gradients whose scale-lengths are comparable with the ion banana width, and includes a magnetic separatrix surface, across which the topological properties of the magnetic field and particle orbits change abruptly. Amore » drift-kinetic particle code XGC0, equipped with a mass-momentum-energy conserving collision operator, is used to study the edge bootstrap current in a realistic diverted magnetic field geometry with a self-consistent radial electric field. When the edge electrons are in the weakly collisional banana regime, surprisingly, the present kinetic simulation confirms that the existing analytic expressions [represented by O. Sauter et al., Phys. Plasmas 6, 2834 (1999)] are still valid in this unconventional region, except in a thin radial layer in contact with the magnetic separatrix. The agreement arises from the dominance of the electron contribution to the bootstrap current compared with ion contribution and from a reasonable separation of the trapped-passing dynamics without a strong collisional mixing. However, when the pedestal electrons are in plateau-collisional regime, there is significant deviation of numerical results from the existing analytic formulas, mainly due to large effective collisionality of the passing and the boundary layer trapped particles in edge region. In a conventional aspect ratio tokamak, the edge bootstrap current from kinetic simulation can be significantly less than that from the Sauter formula if the electron collisionality is high. On the other hand, when the aspect ratio is close to unity, the collisional edge bootstrap current can be significantly greater than that from the Sauter formula. Rapid toroidal rotation of the magnetic field lines at the high field side of a tight aspect-ratio tokamak is believed to be the cause of the different behavior. A new analytic fitting formula, as a simple modification to the Sauter formula, is obtained to bring the analytic expression to a better agreement with the edge kinetic simulation results.« less

Particle acceleration during merging-compression plasma start-up in the Mega Amp Spherical Tokamak

NASA Astrophysics Data System (ADS)

McClements, K. G.; Allen, J. O.; Chapman, S. C.; Dendy, R. O.; Irvine, S. W. A.; Marshall, O.; Robb, D.; Turnyanskiy, M.; Vann, R. G. L.

2018-02-01

Magnetic reconnection occurred during merging-compression plasma start-up in the Mega Amp Spherical Tokamak (MAST), resulting in the prompt acceleration of substantial numbers of ions and electrons to highly suprathermal energies. Accelerated field-aligned ions (deuterons and protons) were detected using a neutral particle analyser at energies up to about 20 keV during merging in early MAST pulses, while nonthermal electrons have been detected indirectly in more recent pulses through microwave bursts. However no increase in soft x-ray emission was observed until later in the merging phase, by which time strong electron heating had been detected through Thomson scattering measurements. A test-particle code CUEBIT is used to model ion acceleration in the presence of an inductive toroidal electric field with a prescribed spatial profile and temporal evolution based on Hall-MHD simulations of the merging process. The simulations yield particle distributions with properties similar to those observed experimentally, including strong field alignment of the fast ions and the acceleration of protons to higher energies than deuterons. Particle-in-cell modelling of a plasma containing a dilute field-aligned suprathermal electron component suggests that at least some of the microwave bursts can be attributed to the anomalous Doppler instability driven by anisotropic fast electrons, which do not produce measurable enhancements in soft x-ray emission either because they are insufficiently energetic or because the nonthermal bremsstrahlung emissivity during this phase of the pulse is below the detection threshold. There is no evidence of runaway electron acceleration during merging, possibly due to the presence of three-dimensional field perturbations.

NIMROD calculations of energetic particle driven toroidal Alfvén eigenmodes

NASA Astrophysics Data System (ADS)

Hou, Yawei; Zhu, Ping; Kim, Charlson C.; Hu, Zhaoqing; Zou, Zhihui; Wang, Zhengxiong; Nimrod Team

2018-01-01

Toroidal Alfvén eigenmodes (TAEs) are gap modes induced by the toroidicity of tokamak plasmas in the absence of continuum damping. They can be excited by energetic particles (EPs) when the EP drive exceeds other dampings, such as electron and ion Landau damping, and collisional and radiative damping. A TAE benchmark case, which was proposed by the International Tokamak Physics Activity group, is studied in this work. The numerical calculations of linear growth of TAEs driven by EPs in a circular-shaped, large aspect ratio tokamak have been performed using the Hybrid Kinetic-MHD (HK-MHD) model implemented in the NIMROD code. This HK-MHD model couples a δf particle-in-cell representation of EPs with the 3D MHD representation of the bulk plasma through moment closure for the momentum conservation equation. Both the excitation of TAEs and their transition to energetic particle modes (EPMs) have been observed. The influence of EP density, temperature, density gradient, and position of the maximum relative density gradient, on the frequency and the growth rate of TAEs are obtained, which are consistent with those from the eigen-analysis calculations, kinetic-MHD, and gyrokinetic simulations for an initial Maxwellian distribution of EPs. The relative pressure gradient of EP at the radial location of the TAE gap, which represents the drive strength of EPs, can strongly affect the growth rate of TAEs. It is demonstrated that the mode transition due to EP drive variation leads to not only the change of frequency but also the change of the mode structure. This mechanism can be helpful in understanding the nonlinear physics of TAE/EPM, such as frequency chirping.

Toroidal Simulations of Sawteeth with Diamagnetic Effects

NASA Astrophysics Data System (ADS)

Beidler, Matthew; Cassak, Paul; Jardin, Stephen

2014-10-01

The sawtooth crash in tokamaks limits the core temperature, adversely impacts confinement, and seeds disruptions. Adequate knowledge of the physics governing the sawtooth crash and a predictive capability of its ramifications has been elusive, including an understanding of incomplete reconnection, i.e., why sawteeth often cease prematurely before processing all available magnetic flux. There is an indication that diamagnetic suppression could play an important role in this phenomenon. While computational tools to study toroidal plasmas have existed for some time, extended-MHD physics have only recently been integrated. Interestingly, incomplete reconnection has been observed in simulations when diamagnetic effects are present. In the current study, we employ the three-dimensional, extended-MHD code M3D-C1 to study the sawtooth crash in a toroidal geometry. In particular, we describe how magnetic reconnection at the q = 1 rational surface evolves when self-consistently increasing diamagnetic effects are present. We also explore how the termination of reconnection may lead to core-relaxing ideal-MHD instabilities.

MHD Calculation of halo currents and vessel forces in NSTX VDEs

NASA Astrophysics Data System (ADS)

Breslau, J. A.; Strauss, H. R.; Paccagnella, R.

2012-10-01

Research tokamaks such as ITER must be designed to tolerate a limited number of disruptions without sustaining significant damage. It is therefore vital to have numerical tools that can accurately predict the effects of these events. The 3D nonlinear extended MHD code M3D [1] can be used to simulate disruptions and calculate the associated wall currents and forces. It has now been validated against halo current data from NSTX experiments in which vertical displacement events (VDEs) were deliberately induced by turning off vertical feedback control. The results of high-resolution numerical simulations at realistic Lundquist numbers show reasonable agreement with the data, supporting a model in which the most dangerously asymmetric currents and heat loads, and the largest horizontal forces, arise in situations where a fast-growing ideal 2,1 external kink mode is destabilized by the scraping-off of flux surfaces with safety factor q>2 during the course of the VDE. [4pt] [1] W. Park, et al., Phys. Plasmas 6 (1999) 1796.

Comparison between measured and predicted turbulence frequency spectra in ITG and TEM regimes

NASA Astrophysics Data System (ADS)

Citrin, J.; Arnichand, H.; Bernardo, J.; Bourdelle, C.; Garbet, X.; Jenko, F.; Hacquin, S.; Pueschel, M. J.; Sabot, R.

2017-06-01

The observation of distinct peaks in tokamak core reflectometry measurements—named quasi-coherent-modes (QCMs)—are identified as a signature of trapped-electron-mode (TEM) turbulence (Arnichand et al 2016 Plasma Phys. Control. Fusion 58 014037). This phenomenon is investigated with detailed linear and nonlinear gyrokinetic simulations using the Gene code. A Tore-Supra density scan is studied, which traverses through a linear (LOC) to saturated (SOC) ohmic confinement transition. The LOC and SOC phases are both simulated separately. In the LOC phase, where QCMs are observed, TEMs are robustly predicted unstable in linear studies. In the later SOC phase, where QCMs are no longer observed, ion-temperature-gradient (ITG) modes are identified. In nonlinear simulations, in the ITG (SOC) phase, a broadband spectrum is seen. In the TEM (LOC) phase, a clear emergence of a peak at the TEM frequencies is seen. This is due to reduced nonlinear frequency broadening of the underlying linear modes in the TEM regime compared with the ITG regime. A synthetic diagnostic of the nonlinearly simulated frequency spectra reproduces the features observed in the reflectometry measurements. These results support the identification of core QCMs as an experimental marker for TEM turbulence.

Development of a fully implicit particle-in-cell scheme for gyrokinetic electromagnetic turbulence simulation in XGC1

NASA Astrophysics Data System (ADS)

Ku, Seung-Hoe; Hager, R.; Chang, C. S.; Chacon, L.; Chen, G.; EPSI Team

2016-10-01

The cancelation problem has been a long-standing issue for long wavelengths modes in electromagnetic gyrokinetic PIC simulations in toroidal geometry. As an attempt of resolving this issue, we implemented a fully implicit time integration scheme in the full-f, gyrokinetic PIC code XGC1. The new scheme - based on the implicit Vlasov-Darwin PIC algorithm by G. Chen and L. Chacon - can potentially resolve cancelation problem. The time advance for the field and the particle equations is space-time-centered, with particle sub-cycling. The resulting system of equations is solved by a Picard iteration solver with fixed-point accelerator. The algorithm is implemented in the parallel velocity formalism instead of the canonical parallel momentum formalism. XGC1 specializes in simulating the tokamak edge plasma with magnetic separatrix geometry. A fully implicit scheme could be a way to accurate and efficient gyrokinetic simulations. We will test if this numerical scheme overcomes the cancelation problem, and reproduces the dispersion relation of Alfven waves and tearing modes in cylindrical geometry. Funded by US DOE FES and ASCR, and computing resources provided by OLCF through ALCC.

A Simulation Model for Drift Resistive Ballooning Turbulence Examining the Influence of Self-consistent Zonal Flows

NASA Astrophysics Data System (ADS)

Cohen, Bruce; Umansky, Maxim; Joseph, Ilon

2015-11-01

Progress is reported on including self-consistent zonal flows in simulations of drift-resistive ballooning turbulence using the BOUT + + framework. Previous published work addressed the simulation of L-mode edge turbulence in realistic single-null tokamak geometry using the BOUT three-dimensional fluid code that solves Braginskii-based fluid equations. The effects of imposed sheared ExB poloidal rotation were included, with a static radial electric field fitted to experimental data. In new work our goal is to include the self-consistent effects on the radial electric field driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We describe a model for including self-consistent zonal flows and an algorithm for maintaining underlying plasma profiles to enable the simulation of steady-state turbulence. We examine the role of Braginskii viscous forces in providing necessary dissipation when including axisymmetric perturbations. We also report on some of the numerical difficulties associated with including the axisymmetric component of the fluctuating fields. This work was performed under the auspices of the U.S. Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory (LLNL-ABS-674950).

C-Mod MHD stability analysis with LHCD

NASA Astrophysics Data System (ADS)

Ebrahimi, Fatima; Bhattacharjee, A.; Delgado, L.; Scott, S.; Wilson, J. R.; Wallace, G. M.; Shiraiwa, S.; Mumgaard, R. T.

2016-10-01

In lower hybrid current drive (LHCD) experiments on the Alcator C-Mod, sawtooth activity could be suppressed as the safety factor q on axis is raised above unity. However, in some of these experiments, after applying LHCD, the onset of MHD mode activity caused the current drive efficiency to significantly drop. Here, we study the stability of these experiments by performing MHD simulations using the NIMROD code starting with experimental EFIT equilibria. First, consistent with the LHCD experiment with no signature of MHD activity, MHD mode activity was also absent in the simulations. Second, for experiments with MHD mode activity, we find that a core n=1 reconnecting mode with dominate poloidal modes of m=2,3 is unstable. This mode is a resistive current-driven mode as its growth rate scales with a negative power of the Lundquist number in the simulations. In addition, with further enhanced reversed-shear q profile in the simulations, a core double tearing mode is found to be unstable. This work is supported by U.S. DOE cooperative agreement DE-FC02-99ER54512 using the Alcator C-Mod tokamak, a DOE Office of Science user facility.

Tomographic reconstruction of tokamak plasma light emission using wavelet-vaguelette decomposition

NASA Astrophysics Data System (ADS)

Schneider, Kai; Nguyen van Yen, Romain; Fedorczak, Nicolas; Brochard, Frederic; Bonhomme, Gerard; Farge, Marie; Monier-Garbet, Pascale

2012-10-01

Images acquired by cameras installed in tokamaks are difficult to interpret because the three-dimensional structure of the plasma is flattened in a non-trivial way. Nevertheless, taking advantage of the slow variation of the fluctuations along magnetic field lines, the optical transformation may be approximated by a generalized Abel transform, for which we proposed in Nguyen van yen et al., Nucl. Fus., 52 (2012) 013005, an inversion technique based on the wavelet-vaguelette decomposition. After validation of the new method using an academic test case and numerical data obtained with the Tokam 2D code, we present an application to an experimental movie obtained in the tokamak Tore Supra. A comparison with a classical regularization technique for ill-posed inverse problems, the singular value decomposition, allows us to assess the efficiency. The superiority of the wavelet-vaguelette technique is reflected in preserving local features, such as blobs and fronts, in the denoised emissivity map.

Tomographic reconstruction of tokamak plasma light emission from single image using wavelet-vaguelette decomposition

NASA Astrophysics Data System (ADS)

Nguyen van yen, R.; Fedorczak, N.; Brochard, F.; Bonhomme, G.; Schneider, K.; Farge, M.; Monier-Garbet, P.

2012-01-01

Images acquired by cameras installed in tokamaks are difficult to interpret because the three-dimensional structure of the plasma is flattened in a non-trivial way. Nevertheless, taking advantage of the slow variation of the fluctuations along magnetic field lines, the optical transformation may be approximated by a generalized Abel transform, for which we propose an inversion technique based on the wavelet-vaguelette decomposition. After validation of the new method using an academic test case and numerical data obtained with the Tokam 2D code, we present an application to an experimental movie obtained in the tokamak Tore Supra. A comparison with a classical regularization technique for ill-posed inverse problems, the singular value decomposition, allows us to assess the efficiency. The superiority of the wavelet-vaguelette technique is reflected in preserving local features, such as blobs and fronts, in the denoised emissivity map.

Numerical verification of bounce-harmonic resonances in neoclassical toroidal viscosity for tokamaks.

PubMed

Kim, Kimin; Park, Jong-Kyu; Boozer, Allen H

2013-05-03

This Letter presents the first numerical verification for the bounce-harmonic (BH) resonance phenomena of the neoclassical transport in a tokamak perturbed by nonaxisymmetric magnetic fields. The BH resonances were predicted by analytic theories of neoclassical toroidal viscosity (NTV), as the parallel and perpendicular drift motions can be resonant and result in a great enhancement of the radial momentum transport. A new drift-kinetic δf guiding-center particle code, POCA, clearly verified that the perpendicular drift motions can reduce the transport by phase-mixing, but in the BH resonances the motions can form closed orbits and particles radially drift out fast. The POCA calculations on resulting NTV torque are largely consistent with analytic calculations, and show that the BH resonances can easily dominate the NTV torque when a plasma rotates in the perturbed tokamak and therefore, is a critical physics for predicting the rotation and stability in the International Thermonuclear Experimental Reactor.

Fluid simulation of tokamak ion temperature gradient turbulence with zonal flow closure model

NASA Astrophysics Data System (ADS)

Yamagishi, Osamu; Sugama, Hideo

2016-03-01

Nonlinear fluid simulation of turbulence driven by ion temperature gradient modes in the tokamak fluxtube configuration is performed by combining two different closure models. One model is a gyrofluid model by Beer and Hammett [Phys. Plasmas 3, 4046 (1996)], and the other is a closure model to reproduce the kinetic zonal flow response [Sugama et al., Phys. Plasmas 14, 022502 (2007)]. By including the zonal flow closure, generation of zonal flows, significant reduction in energy transport, reproduction of the gyrokinetic transport level, and nonlinear upshift on the critical value of gradient scale length are observed.

Fluid simulation of tokamak ion temperature gradient turbulence with zonal flow closure model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yamagishi, Osamu, E-mail: yamagisi@nifs.ac.jp; Sugama, Hideo

Nonlinear fluid simulation of turbulence driven by ion temperature gradient modes in the tokamak fluxtube configuration is performed by combining two different closure models. One model is a gyrofluid model by Beer and Hammett [Phys. Plasmas 3, 4046 (1996)], and the other is a closure model to reproduce the kinetic zonal flow response [Sugama et al., Phys. Plasmas 14, 022502 (2007)]. By including the zonal flow closure, generation of zonal flows, significant reduction in energy transport, reproduction of the gyrokinetic transport level, and nonlinear upshift on the critical value of gradient scale length are observed.

Modeling of the EAST ICRF antenna with ICANT Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qin Chengming; Zhao Yanping; Colas, L.

2007-09-28

A Resonant Double Loop (RDL) antenna for ion-cyclotron range of frequencies (ICRF) on Experimental Advanced Superconducting Tokamak (EAST) is under construction. The new antenna is analyzed using the antenna coupling code ICANT which self-consistently determines the surface currents on all antenna parts. In this work, the modeling of the new ICRF antenna using this code is to assess the near-fields in front of the antenna and analysis its coupling capabilities. Moreover, the antenna reactive radiated power computed by ICANT and shows a good agreement with deduced from Transmission Line (TL) theory.

Modeling of the EAST ICRF antenna with ICANT Code

NASA Astrophysics Data System (ADS)

Qin, Chengming; Zhao, Yanping; Colas, L.; Heuraux, S.

2007-09-01

A Resonant Double Loop (RDL) antenna for ion-cyclotron range of frequencies (ICRF) on Experimental Advanced Superconducting Tokamak (EAST) is under construction. The new antenna is analyzed using the antenna coupling code ICANT which self-consistently determines the surface currents on all antenna parts. In this work, the modeling of the new ICRF antenna using this code is to assess the near-fields in front of the antenna and analysis its coupling capabilities. Moreover, the antenna reactive radiated power computed by ICANT and shows a good agreement with deduced from Transmission Line (TL) theory.

Helical flow in RFX-mod tokamak plasmas

NASA Astrophysics Data System (ADS)

Piron, L.; Zaniol, B.; Bonfiglio, D.; Carraro, L.; Kirk, A.; Marrelli, L.; Martin, R.; Piron, C.; Piovesan, P.; Zuin, M.

2017-05-01

This work presents the first evidence of helical flow in RFX-mod q(a)  <  2 tokamak plasmas. The flow pattern is characterized by the presence of convective cells with m  =  1 and n  =  1 periodicity in the poloidal and toroidal directions, respectively. A similar helical flow deformation has been observed in the same device when operated as a reversed field pinch (RFP). In RFP plasmas, the flow dynamic is tailored by the innermost resonant m  =  1, n  =  7 tearing mode, which sustains the magnetic field configuration through the dynamo mechanism (Bonomo et al 2011 Nucl. Fusion 51 123007). By contrast, in the tokamak experiments presented here, it is strongly correlated with the m  =  1, n  =  1 MHD activity. A helical deformation of the flow pattern, associated with the deformation of the magnetic flux surfaces, is predicted by several codes, such as Specyl (Bonfiglio et al 2005 Phys. Rev. Lett. 94 145001), PIXIE3D (Chacón et al 2008 Phys. Plasmas 15 056103), NIMROD (King et al 2012 Phys. Plasmas 19 055905) and M3D-C1 (Jardin et al 2015 Phys. Rev. Lett. 115 215001). Among them, the 3D fully non-linear PIXIE3D has been used to calculate synthetic flow measurements, using a 2D flow modelling code. Inputs to the code are the PIXIE3D flow maps, the ion emission profiles as calculated by a 1D collisional radiative impurity transport code (Carraro et al 2000 Plasma Phys. Control. Fusion 42 731) and a synthetic diagnostic with the same geometry installed in RFX-mod. Good agreement between the synthetic and the experimental flow behaviour has been obtained, confirming that the flow oscillations observed with the associated convective cells are a signature of helical flow.

Design, simulation and construction of the Taban tokamak

NASA Astrophysics Data System (ADS)

H, R. MIRZAEI; R, AMROLLAHI

2018-04-01

This paper describes the design and construction of the Taban tokamak, which is located in Amirkabir University of Technology, Tehran, Iran. The Taban tokamak was designed for plasma investigation. The design, simulation and construction of essential parts of the Taban tokamak such as the toroidal field (TF) system, ohmic heating (OH) system and equilibrium field system and their power supplies are presented. For the Taban tokamak, the toroidal magnetic coil was designed to produce a maximum field of 0.7 T at R = 0.45 m. The power supply of the TF was a 130 kJ, 0–10 kV capacitor bank. Ripples of toroidal magnetic field at the plasma edge and plasma center are 0.2% and 0.014%, respectively. For the OH system with 3 kA current, the stray field in the plasma region is less than 40 G over 80% of the plasma volume. The power supply of the OH system consists of two stages, as follows. The fast bank stage is a 120 μF, 0–5 kV capacitor that produces 2.5 kA in 400 μs and the slow bank stage is 93 mF, 600 V that can produce a maximum of 3 kA. The equilibrium system can produce uniform magnetic field at plasma volume. This system’s power supply, like the OH system, consists of two stages, so that the fast bank stage is 500 μF, 800 V and the slow bank stage is 110 mF, 200 V.

Using AORSA to simulate helicon waves in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lau, C., E-mail: lauch@ornl.gov; Blazevski, D.; Green, D. L.

2015-12-10

Recent efforts have shown that helicon waves (fast waves at > 20ω{sub ci}) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored,more » it will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects.« less

Using AORSA to simulate helicon waves in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lau, Cornwall H; Jaeger, E. F.; Bertelli, Nicola

2015-01-01

Recent efforts have shown that helicon waves (fast waves at >20 omega(ci)) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored, itmore » will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects.« less

Installation and Testing of ITER Integrated Modeling and Analysis Suite (IMAS) on DIII-D

NASA Astrophysics Data System (ADS)

Lao, L.; Kostuk, M.; Meneghini, O.; Smith, S.; Staebler, G.; Kalling, R.; Pinches, S.

2017-10-01

A critical objective of the ITER Integrated Modeling Program is the development of IMAS to support ITER plasma operation and research activities. An IMAS framework has been established based on the earlier work carried out within the EU. It consists of a physics data model and a workflow engine. The data model is capable of representing both simulation and experimental data and is applicable to ITER and other devices. IMAS has been successfully installed on a local DIII-D server using a flexible installer capable of managing the core data access tools (Access Layer and Data Dictionary) and optionally the Kepler workflow engine and coupling tools. A general adaptor for OMFIT (a workflow engine) is being built for adaptation of any analysis code to IMAS using a new IMAS universal access layer (UAL) interface developed from an existing OMFIT EU Integrated Tokamak Modeling UAL. Ongoing work includes development of a general adaptor for EFIT and TGLF based on this new UAL that can be readily extended for other physics codes within OMFIT. Work supported by US DOE under DE-FC02-04ER54698.

Modeling of RF/MHD coupling using NIMROD and GENRAY

NASA Astrophysics Data System (ADS)

Jenkins, Thomas G.; Schnack, D. D.; Sovinec, C. R.; Hegna, C. C.; Callen, J. D.; Ebrahimi, F.; Kruger, S. E.; Carlsson, J.; Held, E. D.; Ji, J.-Y.; Harvey, R. W.; Smirnov, A. P.

2008-11-01

We summarize ongoing theoretical/numerical work relevant to the development of a self--consistent framework for the inclusion of RF effects in fluid simulations, specifically considering the stabilization of resistive tearing modes in tokamak (DIII--D--like) geometry by electron cyclotron current drive. Previous investigations [T. G. Jenkins et al., Bull. APS 52, 131 (2007)] have demonstrated that relatively simple (though non--self--consistent) models for the RF--induced currents can be incorporated into the fluid equations, and that these currents can markedly reduce the width of the nonlinearly saturated magnetic islands generated by tearing modes. We report our progress toward the self--consistent modeling of these RF--induced currents. The initial interfacing of the NIMROD* code with the GENRAY/CQL3D** codes (which calculate RF propagation and energy/momentum deposition) is explained, equilibration of RF--induced currents over the plasma flux surfaces is investigated, and initial studies exploring the efficient reduction of saturated island widths through time modulation of the ECCD are presented. Conducted as part of the SWIM*** project; funded by U. S. DoE. *www.nimrodteam.org **www.compxco.com ***www.cswim.org

Integrated tokamak modeling: when physics informs engineering and research planning

NASA Astrophysics Data System (ADS)

Poli, Francesca

2017-10-01

Simulations that integrate virtually all the relevant engineering and physics aspects of a real tokamak experiment are a power tool for experimental interpretation, model validation and planning for both present and future devices. This tutorial will guide through the building blocks of an ``integrated'' tokamak simulation, such as magnetic flux diffusion, thermal, momentum and particle transport, external heating and current drive sources, wall particle sources and sinks. Emphasis is given to the connection and interplay between external actuators and plasma response, between the slow time scales of the current diffusion and the fast time scales of transport, and how reduced and high-fidelity models can contribute to simulate a whole device. To illustrate the potential and limitations of integrated tokamak modeling for discharge prediction, a helium plasma scenario for the ITER pre-nuclear phase is taken as an example. This scenario presents challenges because it requires core-edge integration and advanced models for interaction between waves and fast-ions, which are subject to a limited experimental database for validation and guidance. Starting from a scenario obtained by re-scaling parameters from the demonstration inductive ``ITER baseline'', it is shown how self-consistent simulations that encompass both core and edge plasma regions, as well as high-fidelity heating and current drive source models are needed to set constraints on the density, magnetic field and heating scheme. This tutorial aims at demonstrating how integrated modeling, when used with adequate level of criticism, can not only support design of operational scenarios, but also help to asses the limitations and gaps in the available models, thus indicating where improved modeling tools are required and how present experiments can help their validation and inform research planning. Work supported by DOE under DE-AC02-09CH1146.

Gyrokinetic Simulations with External Resonant Magnetic Perturbations: Island Torque and Nonambipolar Transport with Rotation

NASA Astrophysics Data System (ADS)

Waltz, R. E.; Waelbroeck, F. L.

2012-03-01

Static external resonant magnetic perturbations (RMPs) have been added to the δf gyrokinetic code GYRO. This allows nonlinear gyrokinetic simulations of the nonambipolar radial current flow jr and the corresponding plasma torque (density) R[jrBθ/c], induced by islands that break the toroidal symmetry of a tokamak. This extends previous GYRO simulations for the transport of toroidal angular momentum (TAM) [1,2]. The focus is on full torus radial slice electrostatic simulations of induced q=m/n=6/3 islands with widths 5% of the minor radius. The island torque scales with the radial electric field Er the island width w, and the intensity I of the high-n micro-turbulence, as wErI^1/2. The net island torque is null at zero Er rather than at zero toroidal rotation. This means that there is a small co-directed magnetic acceleration to the small diamagnetic co-rotation corresponding to the zero Er which can be called the residual stress [2] from an externally induced island. Finite-beta GYRO simulations of a core radial slice demonstrate island unlocking and the RMP screening. 6pt[1] R.E. Waltz, et al., Phys. Plasmas 14, 122507 (2007). [2] R.E. Waltz, et al., Phys. Plasmas 18, 042504 (2011).

Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment

NASA Astrophysics Data System (ADS)

Boyle, Dennis Patrick

This thesis presents new measurements of core impurity concentrations and transport in plasmas with lithium coatings on all-metal plasma facing components (PFCs) in the Lithium Tokamak Experiment (LTX). LTX is a modest-sized spherical tokamak uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma (as opposed to just the divertor or limiter region in other devices). Lithium (Li) wall-coatings have improved plasma performance and confinement in several tokamaks with carbon (C) PFCs, including the National Spherical Torus Experiment (NSTX). In NSTX, contamination of the core plasma with Li impurities was very low (<0.1%) despite extensive divertor coatings. Low Li levels in NSTX were found to be largely due to neoclassical forces from the high level of C impurities. Studying impurity levels and transport with Li coatings on stainless steel surfaces in LTX is relevant to future devices (including future enhancements to NSTX-Upgrade) with all-metal PFCs. The new measurements in this thesis were enabled by a refurbished Thomson scattering system and improved impurity spectroscopy, primarily using a novel visible spectrometer monitoring several Li, C, and oxygen (O) emission lines. A simple model was used to account for impurities in unmeasured charge states, assuming constant density in the plasma core and constant concentration in the edge. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with 2-4% Li, 0.6-2% C, 0.4-0.7% O, and Z eff<1.2. Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, unlike in NSTX, where collisions with C dominated. Furthermore, neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two, in contrast to NSTX where they differed by an order of magnitude. However, time-independent simulations with MIST indicated that unlike NSTX, neoclassical theory did not fully capture the impurity transport and anomalous transport likely played a significant role in determining impurity profiles.

Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyle, Dennis Patrick

This thesis presents new measurements of core impurity concentrations and transport in plasmas with lithium coatings on all-metal plasma facing components (PFCs) in the Lithium Tokamak Experiment (LTX). LTX is a modest-sized spherical tokamak uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma (as opposed to just the divertor or limiter region in other devices). Lithium (Li) wall-coatings have improved plasma performance and confinement in several tokamaks with carbon (C) PFCs, including the National Spherical Torus Experiment (NSTX). In NSTX, contamination of the core plasma with Li impurities was very low (<0.1%)more » despite extensive divertor coatings. Low Li levels in NSTX were found to be largely due to neoclassical forces from the high level of C impurities. Studying impurity levels and transport with Li coatings on stainless steel surfaces in LTX is relevant to future devices (including future enhancements to NSTX-Upgrade) with all-metal PFCs. The new measurements in this thesis were enabled by a refurbished Thomson scattering system and improved impurity spectroscopy, primarily using a novel visible spectrometer monitoring several Li, C, and oxygen (O) emission lines. A simple model was used to account for impurities in unmeasured charge states, assuming constant density in the plasma core and constant concentration in the edge. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with~2-4% Li, ~0.6-2% C, ~0.4-0.7% O, and Z_eff<1.2. Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, unlike in NSTX, where collisions with C dominated. Furthermore, neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two, in contrast to NSTX where they differed by an order of magnitude. However, time-independent simulations with MIST indicated that unlike NSTX, neoclassical theory did not fully capture the impurity transport and anomalous transport likely played a significant role in determining impurity profiles.« less

Effects of Density and Impurity on Edge Localized Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Zhu, Ping

2017-10-01

Plasma density and impurity concentration are believed to be two of the key elements governing the edge tokamak plasma conditions. Optimal levels of plasma density and impurity concentration in the edge region have been searched for in order to achieve the desired fusion gain and divertor heat/particle load mitigation. However, how plasma density or impurity would affect the edge pedestal stability may have not been well known. Our recent MHD theory modeling and simulations using the NIMROD code have found novel effects of density and impurity on the dynamics of edge-localized modes (ELMs) in tokamaks. First, previous MHD analyses often predict merely a weak stabilizing effect of toroidal flow on ELMs in experimentally relevant regimes. We find that the stabilizing effects on the high- n ELMs from toroidal flow can be significantly enhanced with the increased edge plasma density. Here n denotes the toroidal mode number. Second, the stabilizing effects of the enhanced edge resistivity due to lithium-conditioning on the low- n ELMs in the high confinement (H-mode) discharges in NSTX have been identified. Linear stability analysis of the experimentally constrained equilibrium suggests that the change in the equilibrium plasma density and pressure profiles alone due to lithium-conditioning may not be sufficient for a complete suppression of the low- n ELMs. The enhanced resistivity due to the increased effective electric charge number Zeff after lithium-conditioning provides additional stabilization of the low- n ELMs. These new effects revealed in our theory analyses may help further understand recent ELM experiments and suggest new control schemes for ELM suppression and mitigation in future experiments. They may also pose additional constraints on the optimal levels of plasma density and impurity concentration in the edge region for H-mode tokamak operation. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB101004, the 100 Talent Program of the Chinese Academy of Sciences, and U.S. Department of Energy Grants DE-FG02-86ER53218 and DE-FC02-08ER54975.

Theory and computation of general force balance in non-axisymmetric tokamak equilibria

NASA Astrophysics Data System (ADS)

Park, Jong-Kyu; Logan, Nikolas; Wang, Zhirui; Kim, Kimin; Boozer, Allen; Liu, Yueqiang; Menard, Jonathan

2014-10-01

Non-axisymmetric equilibria in tokamaks can be effectively described by linearized force balance. In addition to the conventional isotropic pressure force, there are three important components that can strongly contribute to the force balance; rotational, anisotropic tensor pressure, and externally given forces, i.e. ∇ --> p + ρv-> . ∇ --> v-> + ∇ --> . <-->Π + f-> = j-> × B-> , especially in, but not limited to, high β and rotating plasmas. Within the assumption of nested flux surfaces, Maxwell equations and energy minimization lead to the modified-generalized Newcomb equation for radial displacements with simple algebraic relations for perpendicular and parallel displacements, including an inhomogeneous term if any of the forces are not explicitly dependent on displacements. The general perturbed equilibrium code (GPEC) solves this force balance consistent with energy and torque given by external perturbations. Local and global behaviors of solutions will be discussed when ∇ --> . <-->Π is solved by the semi-analytic code PENT and will be compared with MARS-K. Any first-principle transport code calculating ∇ --> . <-->Π or f-> , e.g. POCA, can also be incorporated without demanding iterations. This work was supported by DOE Contract DE-AC02-09CH11466.

Compressional Alfvén eigenmodes in rotating spherical tokamak plasmas

DOE PAGES

Smith, H. M.; Fredrickson, E. D.

2017-02-07

Spherical tokamaks often have a considerable toroidal plasma rotation of several tens of kHz. Compressional Alfvén eigenmodes in such devices therefore experience a frequency shift, which if the plasma were rotating as a rigid body, would be a simple Doppler shift. However, since the rotation frequency depends on minor radius, the eigenmodes are affected in a more complicated way. The eigenmode solver CAE3B (Smith et al 2009 Plasma Phys. Control. Fusion 51 075001) has been extended to account for toroidal plasma rotation. The results show that the eigenfrequency shift due to rotation can be approximated by a rigid body rotationmore » with a frequency computed from a spatial average of the real rotation profile weighted with the eigenmode amplitude. To investigate the effect of extending the computational domain to the vessel wall, a simplified eigenmode equation, yet retaining plasma rotation, is solved by a modified version of the CAE code used in Fredrickson et al (2013 Phys. Plasmas 20 042112). Lastly, both solving the full eigenmode equation, as in the CAE3B code, and placing the boundary at the vessel wall, as in the CAE code, significantly influences the calculated eigenfrequencies.« less

Analysis of a tungsten sputtering experiment in DIII-D and code/data validation of high redeposition/reduced erosion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wampler, William R.; Brooks, J. N.; Elder, J. D.

2015-03-29

We analyze a DIII-D tokamak experiment where two tungsten spots on the removable DiMES divertor probe were exposed to 12 s of attached plasma conditions, with moderate strike point temperature and density (~20 eV, ~4.5 × 10 19 m –3), and 3% carbon impurity content. Both very small (1 mm diameter) and small (1 cm diameter) deposited samples were used for assessing gross and net tungsten sputtering erosion. The analysis uses a 3-D erosion/redeposition code package (REDEP/WBC), with input from a diagnostic-calibrated near-surface plasma code (OEDGE), and with focus on charge state resolved impinging carbon ion flux and energy. Themore » tungsten surfaces are primarily sputtered by the carbon, in charge states +1 to +4. We predict high redeposition (~75%) of sputtered tungsten on the 1 cm spot—with consequent reduced net erosion—and this agrees well with post-exposure DiMES probe RBS analysis data. As a result, this study and recent related work is encouraging for erosion lifetime and non-contamination performance of tokamak reactor high-Z plasma facing components.« less

Nonlinear 3D MHD verification study: SpeCyl and PIXIE3D codes for RFP and Tokamak plasmas

NASA Astrophysics Data System (ADS)

Bonfiglio, D.; Cappello, S.; Chacon, L.

2010-11-01

A strong emphasis is presently placed in the fusion community on reaching predictive capability of computational models. An essential requirement of such endeavor is the process of assessing the mathematical correctness of computational tools, termed verification [1]. We present here a successful nonlinear cross-benchmark verification study between the 3D nonlinear MHD codes SpeCyl [2] and PIXIE3D [3]. Excellent quantitative agreement is obtained in both 2D and 3D nonlinear visco-resistive dynamics for reversed-field pinch (RFP) and tokamak configurations [4]. RFP dynamics, in particular, lends itself as an ideal non trivial test-bed for 3D nonlinear verification. Perspectives for future application of the fully-implicit parallel code PIXIE3D to RFP physics, in particular to address open issues on RFP helical self-organization, will be provided. [4pt] [1] M. Greenwald, Phys. Plasmas 17, 058101 (2010) [0pt] [2] S. Cappello and D. Biskamp, Nucl. Fusion 36, 571 (1996) [0pt] [3] L. Chac'on, Phys. Plasmas 15, 056103 (2008) [0pt] [4] D. Bonfiglio, L. Chac'on and S. Cappello, Phys. Plasmas 17 (2010)

Neutron field measurement at the Experimental Advanced Superconducting Tokamak using a Bonner sphere spectrometer

NASA Astrophysics Data System (ADS)

Hu, Zhimeng; Zhong, Guoqiang; Ge, Lijian; Du, Tengfei; Peng, Xingyu; Chen, Zhongjing; Xie, Xufei; Yuan, Xi; Zhang, Yimo; Sun, Jiaqi; Fan, Tieshuan; Zhou, Ruijie; Xiao, Min; Li, Kai; Hu, Liqun; Chen, Jun; Zhang, Hui; Gorini, Giuseppe; Nocente, Massimo; Tardocchi, Marco; Li, Xiangqing; Chen, Jinxiang; Zhang, Guohui

2018-07-01

The neutron field measurement was performed in the Experimental Advanced Superconducting Tokamak (EAST) experimental hall using a Bonner sphere spectrometer (BSS) based on a 3He thermal neutron counter. The measured spectra and the corresponding integrated neutron fluence and dose values deduced from the spectra at two exposed positions were compared to the calculated results obtained by a general Monte Carlo code MCNP5, and good agreements were found. The applicability of a homemade dose survey meter installed at EAST was also verified with the comparison of the ambient dose equivalent H*(10) values measured by the meter and BSS.

The rectangular array of magnetic probes on J-TEXT tokamak.

PubMed

Chen, Zhipeng; Li, Fuming; Zhuang, Ge; Jian, Xiang; Zhu, Lizhi

2016-11-01

The rectangular array of magnetic probes system was newly designed and installed in the torus on J-TEXT tokamak to measure the local magnetic fields outside the last closed flux surface at a single toroidal angle. In the implementation, the experimental results agree well with the theoretical results based on the Spool model and three-dimensional numerical finite element model when the vertical field was applied. Furthermore, the measurements were successfully used as the input of EFIT code to conduct the plasma equilibrium reconstruction. The calculated Faraday rotation angle using the EFIT output is in agreement with the measured one from the three-wave polarimeter-interferometer system.

The rectangular array of magnetic probes on J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Chen, Zhipeng; Li, Fuming; Zhuang, Ge; Jian, Xiang; Zhu, Lizhi

2016-11-01

The rectangular array of magnetic probes system was newly designed and installed in the torus on J-TEXT tokamak to measure the local magnetic fields outside the last closed flux surface at a single toroidal angle. In the implementation, the experimental results agree well with the theoretical results based on the Spool model and three-dimensional numerical finite element model when the vertical field was applied. Furthermore, the measurements were successfully used as the input of EFIT code to conduct the plasma equilibrium reconstruction. The calculated Faraday rotation angle using the EFIT output is in agreement with the measured one from the three-wave polarimeter-interferometer system.

Mean flows and blob velocities in scrape-off layer (SOLT) simulations of an L-mode discharge on Alcator C-Mod

DOE PAGES

Russell, D. A.; Myra, J. R.; D'Ippolito, D. A.; ...

2016-06-10

Two-dimensional scrape-off layer turbulence (SOLT) code simulations are compared with an L-mode discharge on the Alcator C-Mod tokamak [M. Greenwald, et al., Phys. Plasmas 21, 110501 (2014)]. Density and temperature profiles for the simulations were obtained by smoothly fitting Thomson scattering and mirror Langmuir probe (MLP) data from the shot. Simulations differing in turbulence intensity were obtained by varying a dissipation parameter. Mean flow profiles and density fluctuation amplitudes are consistent with those measured by MLP in the experiment and with a Fourier space diagnostic designed to measure poloidal phase velocity. Blob velocities in the simulations were determined from themore » correlation function for density fluctuations, as in the analysis of gas-puff-imaging (GPI) blobs in the experiment. In the simulations, it was found that larger blobs moved poloidally with the ExB flow velocity, v E , in the near-SOL, while smaller fluctuations moved with the group velocity of the dominant linear (interchange) mode, v E + 1/2 v di, where v di is the ion diamagnetic drift velocity. Comparisons are made with the measured GPI correlation velocity for the discharge. The saturation mechanisms operative in the simulation of the discharge are also discussed. In conclusion, it is found that neither sheared flow nor pressure gradient modification can be excluded as saturation mechanisms.« less

Optimization study of normal conductor tokamak for commercial neutron source

NASA Astrophysics Data System (ADS)

Fujita, T.; Sakai, R.; Okamoto, A.

2017-05-01

The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

Estimation of sheath potentials in front of ASDEX upgrade ICRF antenna with SSWICH asymptotic code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Křivská, A., E-mail: alena.krivska@rma.ac.be; Bobkov, V.; Jacquot, J.

Multi-megawatt Ion Cyclotron Range of Frequencies (ICRF) heating became problematic in ASDEX Upgrade (AUG) tokamak after coating of ICRF antenna limiters and other plasma facing components by tungsten. Strong impurity influx was indeed produced at levels of injected power markedly lower than in the previous experiments. It is assumed that the impurity production is mainly driven by parallel component of Radio-Frequency (RF) antenna electric near-field E// that is rectified in sheaths. In this contribution we estimate poloidal distribution of sheath Direct Current (DC) potential in front of the ICRF antenna and simulate its relative variations over the parametric scans performedmore » during experiments, trying to reproduce some of the experimental observations. In addition, relative comparison between two types of AUG ICRF antenna configurations, used for experiments in 2014, has been performed. For this purpose we use the Torino Polytechnic Ion Cyclotron Antenna (TOPICA) code and asymptotic version of the Self-consistent Sheaths and Waves for Ion Cyclotron Heating (SSWICH) code. Further, we investigate correlation between amplitudes of the calculated oscillating sheath voltages and the E// fields computed at the lateral side of the antenna box, in relation with a heuristic antenna design strategy at IPP Garching to mitigate RF sheaths.« less

Polarized fusion, its implications, and plans for a proof-of-principle experiment at the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Sandorfi, A. M.; Deur, A.; Lowry, M. M.; Wei, X.; Pace, D.; Eidietis, N.; Hyatt, A.; Jackson, G. L.; Lanctot, M.; Smith, S.; St-John, H.; Miller, G. W.; Zheng, X.; Baylor, L. R.

2015-10-01

The cross section for the primary fusion reaction in a tokamak, D+t --> α +n, would increase by a factor of 1.5 if the fuels were spin polarized parallel to the local field, rather than randomly oriented. Simulations show further gains in reaction rate would accompany this increase in large-scale machines such as ITER, due to increased alpha heating. The potential realization of such benefits rests on the crucial question of the survival of spin polarization for periods comparable to the energy containment time. Despite encouraging calculations, technical challenges in preparing and handling polarized materials have prevented any direct tests. Advances in three areas - polarized material technologies developed for nuclear and particle physics as well as medical imaging, polymer pellets developed for Inertial Confinement, and cryogenic injection guns developed for fueling tokamaks - have matured to the point where a direct in situ measurement is possible using the mirror reaction, D+3He --> α +p. Designs and simulations of a proof-of-principle experiment at the DIII-D tokamak in San Diego will be discussed. Work carried out under US DOE Contract DE-AC05-06OR23177 supporting Jefferson Lab and General Atomics Internal R&D funding.

Macroscopic erosion of divertor and first wall armour in future tokamaks

NASA Astrophysics Data System (ADS)

Würz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

2002-12-01

Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.

Model Development for VDE Computations in NIMROD

NASA Astrophysics Data System (ADS)

Bunkers, K. J.; Sovinec, C. R.

2017-10-01

Vertical displacement events (VDEs) and the disruptions associated with them have potential for causing considerable physical damage to ITER and other tokamak experiments. We report on simulations of generic axisymmetric VDEs and a vertically unstable case from Alcator C-MOD using the NIMROD code. Previous calculations have been done with closures for heat flux and viscous stress. Initial calculations show that halo current width is dependent on temperature boundary conditions, and so transport together with plasma-surface interaction may play a role in determining halo currents in experiments. The behavior of VDEs with Braginskii thermal conductivity and viscosity closures and Spitzer-like resistivity are investigated for both the generic axisymmetric VDE case and the C-MOD case. This effort is supported by the U.S. Dept. of Energy, Award Numbers DE-FG02-06ER54850 and DE-FC02-08ER54975.

Multi-field plasma sandpile model in tokamaks and applications

NASA Astrophysics Data System (ADS)

Peng, X. D.; Xu, J. Q.

2016-08-01

A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.

Test particles dynamics in the JOREK 3D non-linear MHD code and application to electron transport in a disruption simulation

NASA Astrophysics Data System (ADS)

Sommariva, C.; Nardon, E.; Beyer, P.; Hoelzl, M.; Huijsmans, G. T. A.; van Vugt, D.; Contributors, JET

2018-01-01

In order to contribute to the understanding of runaway electron generation mechanisms during tokamak disruptions, a test particle tracker is introduced in the JOREK 3D non-linear MHD code, able to compute both full and guiding center relativistic orbits. Tests of the module show good conservation of the invariants of motion and consistency between full orbit and guiding center solutions. A first application is presented where test electron confinement properties are investigated in a massive gas injection-triggered disruption simulation in JET-like geometry. It is found that electron populations initialised before the thermal quench (TQ) are typically not fully deconfined in spite of the global stochasticity of the magnetic field during the TQ. The fraction of ‘survivors’ decreases from a few tens down to a few tenths of percent as the electron energy varies from 1 keV to 10 MeV. The underlying mechanism for electron ‘survival’ is the prompt reformation of closed magnetic surfaces at the plasma core and, to a smaller extent, the subsequent reappearance of a magnetic surface at the edge. It is also found that electrons are less deconfined at 10 MeV than at 1 MeV, which appears consistent with a phase averaging effect due to orbit shifts at high energy.

Energetic particle modes of q = 1 high-order harmonics in tokamak plasmas with monotonic weak magnetic shear

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ren, Zhen-Zhen; Wang, Feng; Fu, G. Y.

Linear and nonlinear simulations of high-order harmonics q=1 energetic particle modes excited by trapped energetic particles in tokamaks are carried out using kinetic/magnetohydrodynamic hybrid code M3D-K. It is found that with a flat safety factor profile in the core region, the linear growth rate of high-order harmonics (m=n>1) driven by energetic trapped particles can be higher than the m/n=1/1 component. The high m=n>1 modes become more unstable when the pressure of energetic particles becomes higher. Moreover, it is shown that there exist multiple resonant locations satisfying different resonant conditions in the phase space of energetic particles for the high-order harmonicsmore » modes, whereas there is only one precessional resonance for the m/n=1/1 harmonics. The fluid nonlinearity reduces the saturation level of the n=1 component, while it hardly affects those of the high n components, especially the modes with m=n=3,4. The frequency of these modes does not chirp significantly, which is different with the typical fishbone driven by trapped particles. Lastly, in addition, the flattening region of energetic particle distribution due to high-order harmonics excitation is wider than that due to m/n=1/1 component, although the m/n=1/1 component has a higher saturation amplitude.« less

Energetic particle modes of q = 1 high-order harmonics in tokamak plasmas with monotonic weak magnetic shear

DOE PAGES

Ren, Zhen-Zhen; Wang, Feng; Fu, G. Y.; ...

2017-04-24

Linear and nonlinear simulations of high-order harmonics q=1 energetic particle modes excited by trapped energetic particles in tokamaks are carried out using kinetic/magnetohydrodynamic hybrid code M3D-K. It is found that with a flat safety factor profile in the core region, the linear growth rate of high-order harmonics (m=n>1) driven by energetic trapped particles can be higher than the m/n=1/1 component. The high m=n>1 modes become more unstable when the pressure of energetic particles becomes higher. Moreover, it is shown that there exist multiple resonant locations satisfying different resonant conditions in the phase space of energetic particles for the high-order harmonicsmore » modes, whereas there is only one precessional resonance for the m/n=1/1 harmonics. The fluid nonlinearity reduces the saturation level of the n=1 component, while it hardly affects those of the high n components, especially the modes with m=n=3,4. The frequency of these modes does not chirp significantly, which is different with the typical fishbone driven by trapped particles. Lastly, in addition, the flattening region of energetic particle distribution due to high-order harmonics excitation is wider than that due to m/n=1/1 component, although the m/n=1/1 component has a higher saturation amplitude.« less

Analysis of plasma particle and energy fluxes to material surfaces from tokamak edge turbulence simulations

NASA Astrophysics Data System (ADS)

Umansky, M. V.; Cohen, B. I.; Rognlien, T. D.; Boedo, J. A.; Rudakov, D. L.

2012-10-01

Recent BOUT simulations of edge plasma turbulence in L-mode regime in the boundary region of DIII-D tokamak have demonstrated reasonable match with key edge diagnostics [1]. Order-of-magnitude level agreement has been found in the characteristic amplitude, wavenumber, and frequency of turbulent fluctuations, as compared with experimental data from reciprocating edge Langmuir probe and Beam Emission Spectroscopy systems. Owing to this encouraging agreement, output data from these simulations are analyzed to get insights on physical mechanisms and properties of plasma particle and energy fluxes to material surfaces. Of particular interest is plasma turbulence propagating into, or generated in, the far scrape-off layer region where plasma interacts with material walls. Results of statistical analyses of simulated turbulence plasma transport will be presented and physical implications will be discussed. [4pt] [1] B.I. Cohen et al., APS-DPP 2012

Can a Penning ionization discharge simulate the tokamak scrape-off plasma conditions?

NASA Technical Reports Server (NTRS)

Finkenthal, M.; Littman, A.; Stutman, D.; Kovnovich, S.; Mandelbaum, P.; Schwob, J. L.; Bhatia, A. K.

1990-01-01

The tokamak scrape-off (the region between the vacuum vessel wall and the magnetically confined fusion plasma edge), represents a source/sink for the hot fusion plasma. The electron densities and temperatures are in the ranges 10 to the 11th - 10 to the 13th/cu cm and 1-40 eV, respectively (depending on the size, magnetic field intensity and configuration, plasma current, etc). In the work reported, the electron temperature and density have been estimated in a Penning ionization discharge by comparing its spectroscopic emission in the VUV with that predicted by a collisional radiative model. An attempt to directly compare this emission with that of the tokamak edge is briefly described.

Effect of anomalous transport on kinetic simulations of the H-mode pedestal

NASA Astrophysics Data System (ADS)

Bateman, G.; Pankin, A. Y.; Kritz, A. H.; Rafiq, T.; Park, G. Y.; Ku, S.; Chang, C. S.

2009-11-01

The MMM08 and MMM95 Multi-Mode transport models [1,2], are used to investigate the effect of anomalous transport in XGC0 gyrokinetic simulations [3] of tokamak H-mode pedestal growth. Transport models are implemented in XGC0 using the Framework for Modernization and Componentization of Fusion Modules (FMCFM). Anomalous transport is driven by steep temperature and density gradients and is suppressed by high values of flow shear in the pedestal. The radial electric field, used to calculate the flow shear rate, is computed self-consistently in the XGC0 code with the anomalous transport, Lagrangian charged particle dynamics and neutral particle effects. XGC0 simulations are used to provide insight into how thermal and particle transport, together with the sources of heat and charged particles, determine the shape and growth rate of the temperature and density profiles. [1] F.D. Halpern et al., Phys. Plasmas 15 (2008) 065033; J.Weiland et al., Nucl. Fusion 49 (2009) 965933; A.Kritz et al., EPS (2009) [2] G. Bateman, et al, Phys. Plasmas 5 (1998) 1793 [3] C.S. Chang, S. Ku, H. Weitzner, Phys. Plasmas 11 (2004) 2649

Verification of fractional quasilinear renormalization theory using drift-wave turbulence simulations

NASA Astrophysics Data System (ADS)

Newman, D. E.; Sanchez, R.; Carreras, B. A.; van Milligen, B. Ph.

2005-10-01

A very recent renormalization scheme for turbulent transport has been formulated in terms fractional differential operators [1]. In this contribution, we test it against numerous tracer particle transport experiments carried out in simulations of drift-wave turbulence in slab geometry [2]. The simplified geometry allows that simulations be carried out for a sufficiently large number of decorrelation times so that the long-term dynamics captured by these operators can be made apparent. By changing the relative dominance of the polarization and ExB nolinearities artificially, we tune at will the degree of homogeneity and isotropy of the system. Additionally, externally-driven sheared flows can also be considered. This wide spectrum of options creates a superb environment to test the strengths and weaknesses of the fractional renormalization formalism. With it, the potential for application to more realistic geometries such as those in state-of-the-art tokamak turbulence codes will be assessed.References[1] R. S'anchez, B.A. Carreras, D.E. Newman, V. Lynch and B.Ph. van Milligen, submitted (2005) [2] D.E. Newman, P.W. Terry, P.H. Diamond and Y. Liang, Phys. Fluids B 5, 1140 (1993)

Effect of resonant magnetic perturbations on microturbulence in DIII-D pedestal

DOE PAGES

Holod, I.; Lin, Z.; Taimourzadeh, S.; ...

2016-10-03

Vacuum resonant magnetic perturbations (RMP) applied to otherwise axisymmetric tokamak plasmas produce in general a combination of non-resonant effects that preserve closed flux surfaces (kink response) and resonant effects that introduce magnetic islands and/or stochasticity (tearing response). The effect of the plasma kink response on the linear stability and nonlinear transport of edge turbulence is studied using the gyrokinetic toroidal code GTC for a DIII-D plasma with applied n = 2 vacuum RMP. GTC simulations use the 3D equilibrium of DIII-D discharge 158103 (Nazikian et al 2015 Phys. Rev. Lett. 114 105002), which is provided by nonlinear ideal MHD VMECmore » equilibrium solver in order to include the effect of the plasma kink response to the external field but to exclude island formation at rational surfaces. Analysis using the GTC simulation results reveal no increase of growth rates for the electrostatic drift wave instability and for the electromagnetic kinetic-ballooning mode in the presence of the plasma kink response to the RMP. Moreover, nonlinear electrostatic simulations show that the effect of the 3D equilibrium on zonal flow damping is very weak and found to be insufficient to modify turbulent transport in the electrostatic turbulence.« less

Constrained ripple optimization of Tokamak bundle divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hively, L.M.; Rome, J.A.; Lynch, V.E.

1983-02-01

Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ..xi.. B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have lowmore » on-axis ripple (<0.2%) so that, now, most banana-trapped fast ions are confined. Only those ions with banana tips near the outside region (absolute value theta < or equal to 45/sup 0/) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded.« less

Optimal control of a coupled partial and ordinary differential equations system for the assimilation of polarimetry Stokes vector measurements in tokamak free-boundary equilibrium reconstruction with application to ITER

NASA Astrophysics Data System (ADS)

Faugeras, Blaise; Blum, Jacques; Heumann, Holger; Boulbe, Cédric

2017-08-01

The modelization of polarimetry Faraday rotation measurements commonly used in tokamak plasma equilibrium reconstruction codes is an approximation to the Stokes model. This approximation is not valid for the foreseen ITER scenarios where high current and electron density plasma regimes are expected. In this work a method enabling the consistent resolution of the inverse equilibrium reconstruction problem in the framework of non-linear free-boundary equilibrium coupled to the Stokes model equation for polarimetry is provided. Using optimal control theory we derive the optimality system for this inverse problem. A sequential quadratic programming (SQP) method is proposed for its numerical resolution. Numerical experiments with noisy synthetic measurements in the ITER tokamak configuration for two test cases, the second of which is an H-mode plasma, show that the method is efficient and that the accuracy of the identification of the unknown profile functions is improved compared to the use of classical Faraday measurements.

Recent Doppler Backscattering results from EAST tokamak

NASA Astrophysics Data System (ADS)

Zhou, Chu; Liu, Adi; Zhang, Xiaohui; Hu, Jianqiang; Wang, Mingyuan; Yu, Changxuan; Liu, Wandong; Li, Hong; Lan, Tao; Sun, Xuan; Xie, Jinlin; Ding, Weixing; CAS Key Laboratory of Geospace Environment, University of Science and Technology of China Team; Department of Physics and Astronomy, University of California at Los Angeles Collaboration

2013-10-01

A Doppler reflectometer system has recently been installed in the EAST tokamak. It includes two separated systems, one for Q-band and the other for V-band. The optical system consists of a fixed flat mirror and a steerable parabolic mirror, which enabling the measurement of perpendicular wave number in the range of 4-22/cm, with the wave number resolution around 2/cm, while the radial location can cover the whole minor radius for L mode and the whole pedestal for H mode on EAST. A 2D Gaussion Ray tracing code is used to calculate the scattering location, the perpendicular wave number and the resolution. In EAST last experimental campaign the Doppler shifted signals have been obtained and the radial profiles of the perpendicular propagation velocity during L-mode and H-mode are calculated. The Er evolution during L-H and H-L transition have also been measured. The two separated systems are also used as a poloidal coherent system together to study the GAM in EAST tokamak.

Maximum entropy reconstruction of poloidal magnetic field and radial electric field profiles in tokamaks

NASA Astrophysics Data System (ADS)

Chen, Yihang; Xiao, Chijie; Yang, Xiaoyi; Wang, Tianbo; Xu, Tianchao; Yu, Yi; Xu, Min; Wang, Long; Lin, Chen; Wang, Xiaogang

2017-10-01

The Laser-driven Ion beam trace probe (LITP) is a new diagnostic method for measuring poloidal magnetic field (Bp) and radial electric field (Er) in tokamaks. LITP injects a laser-driven ion beam into the tokamak, and Bp and Er profiles can be reconstructed using tomography methods. A reconstruction code has been developed to validate the LITP theory, and both 2D reconstruction of Bp and simultaneous reconstruction of Bp and Er have been attained. To reconstruct from experimental data with noise, Maximum Entropy and Gaussian-Bayesian tomography methods were applied and improved according to the characteristics of the LITP problem. With these improved methods, a reconstruction error level below 15% has been attained with a data noise level of 10%. These methods will be further tested and applied in the following LITP experiments. Supported by the ITER-CHINA program 2015GB120001, CHINA MOST under 2012YQ030142 and National Natural Science Foundation Abstract of China under 11575014 and 11375053.

NIMROD modeling of poloidal flow damping in tokamaks using kinetic closures

NASA Astrophysics Data System (ADS)

Jepson, J. R.; Hegna, C. C.; Held, E. D.

2017-10-01

Calculations of poloidal flow damping in a tokamak are undertaken using two different implementations of the ion drift kinetic equation (DKE) in the extended MHD code NIMROD. The first approach is hybrid fluid/kinetic and uses a Chapman Enskog-like (CEL) Ansatz. Closure of the evolving lower-order fluid moment equations for n, V , and T is provided by solutions to the ion CEL-DKE written in the macroscopic flow reference frame. The second implementation solves the DKE using a delta-f approach. Here, the delta-f distribution describes all of the information beyond a static, lowest-order Maxwellian. We compare the efficiency and accuracy of these two approaches for a simple initial value problem that monitors the relaxation of the poloidal flow profile in high- and low-aspect-ratio tokamak geometry. The computation results are compared against analytic predictions of time dependent closures for the parallel viscous force. Supported by DoE Grants DE-FG02-86ER53218 and DE-FG02-04ER54746.

Full orbit computations of ripple-induced fusion {alpha}-particle losses from burning tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

McClements, K.G.

A full orbit code is used to compute collisionless losses of fusion {alpha} particles from three proposed burning plasma tokamaks: the International Tokamak Experimental Reactor (ITER); a spherical tokamak power plant (STPP) [T. C. Hender, A. Bond, J. Edwards, P. J. Karditsas, K. G. McClements, J. Mustoe, D. V. Sherwood, G. M. Voss, and H. R. Wilson, Fusion Eng. Des. 48, 255 (2000)]; and a spherical tokamak components test facility (CTF) [H. R. Wilson, G. M. Voss, R. J. Akers, L. Appel, A. Dnestrovskij, O. Keating, T. C. Hender, M. J. Hole, G. Huysmans, A. Kirk, P. J. Knight, M.more » Loughlin, K. G. McClements, M. R. O'Brien, and D. Yu. Sychugov, Proceedings of the 20th IAEA Fusion Energy Conference, Invited Paper FT/3-1Ra]. It has been suggested that {alpha} particle transport could be enhanced due to cyclotron resonance with the toroidal magnetic field ripple. However, calculations for inductive operation in ITER yield a loss rate that appears to be broadly consistent with the predictions of guiding center theory, falling monotonically as the number of toroidal field coils N is increased (and hence the ripple amplitude is decreased). For STPP and CTF the loss rate does not decrease monotonically with N, but collisionless losses are generally low in absolute terms. As in the case of ITER, there is no evidence that finite Larmor radius effects would seriously degrade fusion {alpha}-particle confinement.« less

The Role of an Electric Field in the Formation of a Detached Regime in Tokamak Plasma

NASA Astrophysics Data System (ADS)

Senichenkov, I.; Kaveeva, E.; Rozhansky, V.; Sytova, E.; Veselova, I.; Voskoboynikov, S.; Coster, D.

2018-03-01

Modeling of the transition to the detachment of ASDEX Upgrade tokamak plasma with increasing density is performed using the SOLPS-ITER numerical code with a self-consistent account of drifts and currents. Their role in plasma redistribution both in the confinement region and in the scrape-off layer (SOL) is investigated. The mechanism of high field side high-density formation in the SOL in the course of detachment is suggested. In the full detachment regime, when the cold plasma region expands above the X-point and reaches closed magnetic-flux surfaces, plasma perturbation in a confined region may lead to a change in the confinement regime.

Sustained high βN plasmas on EAST tokamak

NASA Astrophysics Data System (ADS)

Gao, Xiang; the EAST team

2018-05-01

Sustained high normalized beta (βN ∼ 1.9) plasmas with an ITER-like tungsten divertor have been achieved on EAST tokamak recently. The high power NBI heating system of 4.8 MW and the 4.6 GHz lower hybrid wave of 1 MW were developed and applied to produce edge and internal transport barriers in high βN discharges. The central flat q profile with q (ρ) ∼ 1 at ρ < 0.3 region and edge safety factor q95 = 4.7 is identified by the multi-channel far-infrared laser polarimeter and the EFIT code. The fraction of non-inductive current is about 40%. The relation between fishbone activity and ITB formation is observed and discussed.

Self-consistent computation of the electric field near ICRH antennas. Application to the Tore Supra antenna

NASA Astrophysics Data System (ADS)

Pécoul, S.; Heuraux, S.; Koch, R.; Leclert, G.; Bécoulet, A.; Colas, L.

1999-09-01

Self-consistent calculations of the 3D electric field patterns between the screen and the plasma have been made with the ICANT code for realistic antennas. Here we explain how the ICRH antennas of the Tore Supra tokamak are modelled.

Self-consistent computation of the electric field near ICRH antennas. Application to the Tore Supra antenna

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pecoul, S.; Heuraux, S.; Koch, R.

1999-09-20

Self-consistent calculations of the 3D electric field patterns between the screen and the plasma have been made with the ICANT code for realistic antennas. Here we explain how the ICRH antennas of the Tore Supra tokamak are modelled.

Non-linear gyrokinetic simulations of microturbulence in TCV electron internal transport barriers

NASA Astrophysics Data System (ADS)

Lapillonne, X.; Brunner, S.; Sauter, O.; Villard, L.; Fable, E.; Görler, T.; Jenko, F.; Merz, F.

2011-05-01

Using the local (flux-tube) version of the Eulerian code GENE (Jenko et al 2000 Phys. Plasmas 7 1904), gyrokinetic simulations of microturbulence were carried out considering parameters relevant to electron-internal transport barriers (e-ITBs) in the TCV tokamak (Sauter et al 2005 Phys. Rev. Lett. 94 105002), generated under conditions of low or negative shear. For typical density and temperature gradients measured in such barriers, the corresponding simulated fluctuation spectra appears to simultaneously contain longer wavelength trapped electron modes (TEMs, for typically k⊥ρi < 0.5, k⊥ being the characteristic perpendicular wavenumber and ρi the ion Larmor radius) and shorter wavelength ion temperature gradient modes (ITG, k⊥ρi > 0.5). The contributions to the electron particle flux from these two types of modes are, respectively, outward/inward and may cancel each other for experimentally realistic gradients. This mechanism may partly explain the feasibility of e-ITBs. The non-linear simulation results confirm the predictions of a previously developed quasi-linear model (Fable et al 2010 Plasma Phys. Control. Fusion 52 015007), namely that the stationary condition of zero particle flux is obtained through the competitive contributions of ITG and TEM. A quantitative comparison of the electron heat flux with experimental estimates is presented as well.

Simulation study of disruption characteristics in KSTAR

NASA Astrophysics Data System (ADS)

Lee, Jongkyu; Kim, J. Y.; Kessel, C. E.; Poli, F.

2012-10-01

A detailed simulation study of disruption in KSTAR had been performed using the Tokamak Simulation Code(TSC) [1] during the initial design phase of KSTAR [2]. Recently, however, a partial modification in the structure of passive plate was made in relation to reduce eddy current and increase the efficiency of control of vertical position. A substantial change can then occur in disruption characteristics and plasma behavior during disruption due to changes in passive plate structure. Because of this, growth rate of vertical instability is expected to be increased and eddy current and its associated electomagnetic force are expected to be reduced. To check this in more detail, a new simulation study is here given with modified passive plate structure of KSTAR. In particular, modeling of vertical disruption that is vertical displacement event (VDE) was carried out. We calculated vertical growth rate for a drift phase of plasma and electromagnetic force acting on PFC structures and compared the results between in a new model and an old model. [4pt] [1] S.C. Jardin, N. Pomphrey and J. Delucia, J. Comp. Phys. 66, 481 (1986).[0pt] [2] J.Y. Kim, S.Y. Cho and KSTAR Team, Disruption load analysis on KSTAR PFC structures, J. Accel. Plasma Res. 5, 149 (2000).

Experiments and Simulations of ITER-like Plasmas in Alcator C-Mod

DOE Office of Scientific and Technical Information (OSTI.GOV)

.R. Wilson, C.E. Kessel, S. Wolfe, I.H. Hutchinson, P. Bonoli, C. Fiore, A.E. Hubbard, J. Hughes, Y. Lin, Y. Ma, D. Mikkelsen, M. Reinke, S. Scott, A.C.C. Sips, S. Wukitch and the C-Mod Team

Alcator C-Mod is performing ITER-like experiments to benchmark and verify projections to 15 MA ELMy H-mode Inductive ITER discharges. The main focus has been on the transient ramp phases. The plasma current in C-Mod is 1.3 MA and toroidal field is 5.4 T. Both Ohmic and ion cyclotron (ICRF) heated discharges are examined. Plasma current rampup experiments have demonstrated that (ICRF and LH) heating in the rise phase can save voltseconds (V-s), as was predicted for ITER by simulations, but showed that the ICRF had no effect on the current profile versus Ohmic discharges. Rampdown experiments show an overcurrent inmore » the Ohmic coil (OH) at the H to L transition, which can be mitigated by remaining in H-mode into the rampdown. Experiments have shown that when the EDA H-mode is preserved well into the rampdown phase, the density and temperature pedestal heights decrease during the plasma current rampdown. Simulations of the full C-Mod discharges have been done with the Tokamak Simulation Code (TSC) and the Coppi-Tang energy transport model is used with modified settings to provide the best fit to the experimental electron temperature profile. Other transport models have been examined also. __________________________________________________« less

Particle-in-cell simulation study of the scaling of asymmetric magnetic reconnection with in-plane flow shear

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doss, C. E.; Cassak, P. A., E-mail: Paul.Cassak@mail.wvu.edu; Swisdak, M.

2016-08-15

We investigate magnetic reconnection in systems simultaneously containing asymmetric (anti-parallel) magnetic fields, asymmetric plasma densities and temperatures, and arbitrary in-plane bulk flow of plasma in the upstream regions. Such configurations are common in the high-latitudes of Earth's magnetopause and in tokamaks. We investigate the convection speed of the X-line, the scaling of the reconnection rate, and the condition for which the flow suppresses reconnection as a function of upstream flow speeds. We use two-dimensional particle-in-cell simulations to capture the mixing of plasma in the outflow regions better than is possible in fluid modeling. We perform simulations with asymmetric magnetic fields,more » simulations with asymmetric densities, and simulations with magnetopause-like parameters where both are asymmetric. For flow speeds below the predicted cutoff velocity, we find good scaling agreement with the theory presented in Doss et al. [J. Geophys. Res. 120, 7748 (2015)]. Applications to planetary magnetospheres, tokamaks, and the solar wind are discussed.« less

Comparison of runaway electron generation parameters in small, medium-sized and large tokamaks—A survey of experiments in COMPASS, TCV, ASDEX-Upgrade and JET

NASA Astrophysics Data System (ADS)

Plyusnin, V. V.; Reux, C.; Kiptily, V. G.; Pautasso, G.; Decker, J.; Papp, G.; Kallenbach, A.; Weinzettl, V.; Mlynar, J.; Coda, S.; Riccardo, V.; Lomas, P.; Jachmich, S.; Shevelev, A. E.; Alper, B.; Khilkevitch, E.; Martin, Y.; Dux, R.; Fuchs, C.; Duval, B.; Brix, M.; Tardini, G.; Maraschek, M.; Treutterer, W.; Giannone, L.; Mlynek, A.; Ficker, O.; Martin, P.; Gerasimov, S.; Potzel, S.; Paprok, R.; McCarthy, P. J.; Imrisek, M.; Boboc, A.; Lackner, K.; Fernandes, A.; Havlicek, J.; Giacomelli, L.; Vlainic, M.; Nocente, M.; Kruezi, U.; COMPASS Team; TCV Team; ASDEX-Upgrade Team; EUROFusion MST1 Team; contributors, JET

2018-01-01

This paper presents a survey of the experiments on runaway electrons (RE) carried out recently in frames of EUROFusion Consortium in different tokamaks: COMPASS, ASDEX-Upgrade, TCV and JET. Massive gas injection (MGI) has been used in different scenarios for RE generation in small and medium-sized tokamaks to elaborate the most efficient and reliable ones for future RE experiments. New data on RE generated at disruptions in COMPASS and ASDEX-Upgrade was collected and added to the JET database. Different accessible parameters of disruptions, such as current quench rate, conversion rate of plasma current into runaways, etc have been analysed for each tokamak and compared to JET data. It was shown, that tokamaks with larger geometrical sizes provide the wider limits for spatial and temporal variation of plasma parameters during disruptions, thus extending the parameter space for RE generation. The second part of experiments was dedicated to study of RE generation in stationary discharges in COMPASS, TCV and JET. Injection of Ne/Ar have been used to mock-up the JET MGI runaway suppression experiments. Secondary RE avalanching was identified and quantified for the first time in the TCV tokamak in RE generating discharges after massive Ne injection. Simulations of the primary RE generation and secondary avalanching dynamics in stationary discharges has demonstrated that RE current fraction created via avalanching could achieve up to 70-75% of the total plasma current in TCV. Relaxations which are reminiscent the phenomena associated to the kinetic instability driven by RE have been detected in RE discharges in TCV. Macroscopic parameters of RE dominating discharges in TCV before and after onset of the instability fit well to the empirical instability criterion, which was established in the early tokamaks and examined by results of recent numerical simulations.

Fast Low-to-High Confinement Mode Bifurcation Dynamics in a Tokamak Edge Plasma Gyrokinetic Simulation.

PubMed

Chang, C S; Ku, S; Tynan, G R; Hager, R; Churchill, R M; Cziegler, I; Greenwald, M; Hubbard, A E; Hughes, J W

2017-04-28

Transport barrier formation and its relation to sheared flows in fluids and plasmas are of fundamental interest in various natural and laboratory observations and of critical importance in achieving an economical energy production in a magnetic fusion device. Here we report the first observation of an edge transport barrier formation event in an electrostatic gyrokinetic simulation carried out in a realistic diverted tokamak edge geometry under strong forcing by a high rate of heat deposition. The results show that turbulent Reynolds-stress-driven sheared E×B flows act in concert with neoclassical orbit loss to quench turbulent transport and form a transport barrier just inside the last closed magnetic flux surface.

Multi-code analysis of scrape-off layer filament dynamics in MAST

NASA Astrophysics Data System (ADS)

Militello, F.; Walkden, N. R.; Farley, T.; Gracias, W. A.; Olsen, J.; Riva, F.; Easy, L.; Fedorczak, N.; Lupelli, I.; Madsen, J.; Nielsen, A. H.; Ricci, P.; Tamain, P.; Young, J.

2016-11-01

Four numerical codes are employed to investigate the dynamics of scrape-off layer filaments in tokamak relevant conditions. Experimental measurements were taken in the MAST device using visual camera imaging, which allows the evaluation of the perpendicular size and velocity of the filaments, as well as the combination of density and temperature associated with the perturbation. A new algorithm based on the light emission integrated along the field lines associated with the position of the filament is developed to ensure that it is properly detected and tracked. The filaments are found to have velocities of the order of 1~\\text{km}~{{\\text{s}}-1} , a perpendicular diameter of around 2-3 cm and a density amplitude 2-3.5 times the background plasma. 3D and 2D numerical codes (the STORM module of BOUT++, GBS, HESEL and TOKAM3X) are used to reproduce the motion of the observed filaments with the purpose of validating the codes and of better understanding the experimental data. Good agreement is found between the 3D codes. The seeded filament simulations are also able to reproduce the dynamics observed in experiments with accuracy up to the experimental errorbar levels. In addition, the numerical results showed that filaments characterised by similar size and light emission intensity can have quite different dynamics if the pressure perturbation is distributed differently between density and temperature components. As an additional benefit, several observations on the dynamics of the filaments in the presence of evolving temperature fields were made and led to a better understanding of the behaviour of these coherent structures.

Simulation of density fluctuations before the L-H transition for Hydrogen and Deuterium plasmas in the DIII-D tokamak using the BOUT++ code

DOE PAGES

Wang, Y. M.; Xu, X. Q.; Yan, Z.; ...

2018-01-05

A six-field two-fluid model has been used to simulate density fluctuations. The equilibrium is generated by experimental measurements for both Deuterium (D) and Hydrogen (H) plasmas at the lowest densities of DIII-D low to high confinement (L-H) transition experiments. In linear simulations, the unstable modes are found to be resistive ballooning modes with the most unstable mode number n=30 ormore » $$k_\\theta\\rho_i\\sim0.12$$ . The ion diamagnetic drift and $$E\\times B$$ convection flow are balanced when the radial electric field (E r) is calculated from the pressure profile without net flow. The curvature drift plays an important role in this stage. Two poloidally counter propagating modes are found in the nonlinear simulation of the D plasma at electron density $$n_e\\sim1.5\\times10^{19}$$ m -3 near the separatrix while a single ion mode is found in the H plasma at the similar lower density, which are consistent with the experimental results measured by the beam emission spectroscopy (BES) diagnostic on the DIII-D tokamak. The frequency of the electron modes and the ion modes are about 40kHz and 10 kHz respectively. The poloidal wave number $$k_\\theta$$ is about 0.2 cm -1 ($$k_\\theta\\rho_i\\sim0.05$$ ) for both ion and electron modes. The particle flux, ion and electron heat fluxes are~3.5–6 times larger for the H plasma than the D plasma, which makes it harder to achieve H-mode for the same heating power. The change of the atomic mass number A from 2 to 1 using D plasma equilibrium make little difference on the flux. Increase the electric field will suppress the density fluctuation. In conclusion, the electric field scan and ion mass scan results show that the dual-mode results primarily from differences in the profiles rather than the ion mass.« less

Simulation of density fluctuations before the L-H transition for Hydrogen and Deuterium plasmas in the DIII-D tokamak using the BOUT++ code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Y. M.; Xu, X. Q.; Yan, Z.

A six-field two-fluid model has been used to simulate density fluctuations. The equilibrium is generated by experimental measurements for both Deuterium (D) and Hydrogen (H) plasmas at the lowest densities of DIII-D low to high confinement (L-H) transition experiments. In linear simulations, the unstable modes are found to be resistive ballooning modes with the most unstable mode number n=30 ormore » $$k_\\theta\\rho_i\\sim0.12$$ . The ion diamagnetic drift and $$E\\times B$$ convection flow are balanced when the radial electric field (E r) is calculated from the pressure profile without net flow. The curvature drift plays an important role in this stage. Two poloidally counter propagating modes are found in the nonlinear simulation of the D plasma at electron density $$n_e\\sim1.5\\times10^{19}$$ m -3 near the separatrix while a single ion mode is found in the H plasma at the similar lower density, which are consistent with the experimental results measured by the beam emission spectroscopy (BES) diagnostic on the DIII-D tokamak. The frequency of the electron modes and the ion modes are about 40kHz and 10 kHz respectively. The poloidal wave number $$k_\\theta$$ is about 0.2 cm -1 ($$k_\\theta\\rho_i\\sim0.05$$ ) for both ion and electron modes. The particle flux, ion and electron heat fluxes are~3.5–6 times larger for the H plasma than the D plasma, which makes it harder to achieve H-mode for the same heating power. The change of the atomic mass number A from 2 to 1 using D plasma equilibrium make little difference on the flux. Increase the electric field will suppress the density fluctuation. In conclusion, the electric field scan and ion mass scan results show that the dual-mode results primarily from differences in the profiles rather than the ion mass.« less

Progress on the development of FullWave, a Hot and Cold Plasma Parallel Full Wave Code

NASA Astrophysics Data System (ADS)

Spencer, J. Andrew; Svidzinski, Vladimir; Zhao, Liangji; Kim, Jin-Soo

2017-10-01

FullWave is being developed at FAR-TECH, Inc. to simulate RF waves in hot inhomogeneous magnetized plasmas without making small orbit approximations. FullWave is based on a meshless formulation in configuration space on non-uniform clouds of computational points (CCP) adapted to better resolve plasma resonances, antenna structures and complex boundaries. The linear frequency domain wave equation is formulated using two approaches: for cold plasmas the local cold plasma dielectric tensor is used (resolving resonances by particle collisions), while for hot plasmas the conductivity kernel is calculated. The details of FullWave and some preliminary results will be presented, including: 1) a monitor function based on analytic solutions of the cold-plasma dispersion relation; 2) an adaptive CCP based on the monitor function; 3) construction of the finite differences for approximation of derivatives on adaptive CCP; 4) results of 2-D full wave simulations in the cold plasma model in tokamak geometry using the formulated approach for ECRH, ICRH and Lower Hybrid range of frequencies. Work is supported by the U.S. DOE SBIR program.

Numerical simulation of the multiple core localized low shear toroidal Alfvenic eigenmodes

NASA Astrophysics Data System (ADS)

Wang, Wenjia; Zhou, Deng; Hu, Youjun; Ming, Yue

2018-03-01

In modern tokamak experiments, scenarios with weak central magnetic shear has been proposed. It is necessary to study the Alfvenic mode activities in such scenarios. Theoretical researches have predicted the multiplicity of core-localized toroidally induced Alfvenic eigenmodes for ɛ/s > 1, where ɛ is the inverse aspect ratio and s is magnetic shear. We numerically investigate the existence of multiplicity of core-localized TAEs and mode characteristics using NOVA code in the present work. We firstly verify the existence of the multiplicity for zero beta plasma and the even mode at the forbidden zone. For finite beta plasma, the mode parities become more distinguishable, and the frequencies of odd modes are close to the upper tip of the continuum, while the frequencies of even modes are close to the lower tip of the continuum. Their frequencies are well separated by the forbidden zone. With the increasing value of ɛ/s, more modes with multiple radial nodes will appear, which is in agreement with theoretical prediction. The discrepancy between theoretical prediction and our numerical simulation is also discussed in the main text.

Real-time diamagnetic flux measurements on ASDEX Upgrade.

PubMed

Giannone, L; Geiger, B; Bilato, R; Maraschek, M; Odstrčil, T; Fischer, R; Fuchs, J C; McCarthy, P J; Mertens, V; Schuhbeck, K H

2016-05-01

Real-time diamagnetic flux measurements are now available on ASDEX Upgrade. In contrast to the majority of diamagnetic flux measurements on other tokamaks, no analog summation of signals is necessary for measuring the change in toroidal flux or for removing contributions arising from unwanted coupling to the plasma and poloidal field coil currents. To achieve the highest possible sensitivity, the diamagnetic measurement and compensation coil integrators are triggered shortly before plasma initiation when the toroidal field coil current is close to its maximum. In this way, the integration time can be chosen to measure only the small changes in flux due to the presence of plasma. Two identical plasma discharges with positive and negative magnetic field have shown that the alignment error with respect to the plasma current is negligible. The measured diamagnetic flux is compared to that predicted by TRANSP simulations. The poloidal beta inferred from the diamagnetic flux measurement is compared to the values calculated from magnetic equilibrium reconstruction codes. The diamagnetic flux measurement and TRANSP simulation can be used together to estimate the coupled power in discharges with dominant ion cyclotron resonance heating.

Self-Organized Stationary States of Tokamaks

DOE PAGES

Jardin, S. C.; Ferraro, N.; Krebs, I.

2015-11-17

We demonstrate that in a 3D resistive magnetohydrodynamic (MHD) simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to non-linearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary non-sawtoothing “hybrid” discharges, often referred to as “flux-pumping”.

Neural network evaluation of tokamak current profiles for real time control

NASA Astrophysics Data System (ADS)

Wróblewski, Dariusz

1997-02-01

Active feedback control of the current profile, requiring real-time determination of the current profile parameters, is envisioned for tokamaks operating in enhanced confinement regimes. The distribution of toroidal current in a tokamak is now routinely evaluated based on external (magnetic probes, flux loops) and internal (motional Stark effect) measurements of the poloidal magnetic field. However, the analysis involves reconstruction of magnetohydrodynamic equilibrium and is too intensive computationally to be performed in real time. In the present study, a neural network is used to provide a mapping from the magnetic measurements (internal and external) to selected parameters of the safety factor profile. The single-pass, feedforward calculation of output of a trained neural network is very fast, making this approach particularly suitable for real-time applications. The network was trained on a large set of simulated equilibrium data for the DIII-D tokamak. The database encompasses a large variety of current profiles including the hollow current profiles important for reversed central shear operation. The parameters of safety factor profile (a quantity related to the current profile through the magnetic field tilt angle) estimated by the neural network include central safety factor, q0, minimum value of q, qmin, and the location of qmin. Very good performance of the trained neural network both for simulated test data and for experimental datais demonstrated.

Neural network evaluation of tokamak current profiles for real time control (abstract)

NASA Astrophysics Data System (ADS)

Wróblewski, Dariusz

1997-01-01

Active feedback control of the current profile, requiring real-time determination of the current profile parameters, is envisioned for tokamaks operating in enhanced confinement regimes. The distribution of toroidal current in a tokamak is now routinely evaluated based on external (magnetic probes, flux loops) and internal (motional Stark effect) measurements of the poloidal magnetic field. However, the analysis involves reconstruction of magnetohydrodynamic equilibrium and is too intensive computationally to be performed in real time. In the present study, a neural network is used to provide a mapping from the magnetic measurements (internal and external) to selected parameters of the safety factor profile. The single-pass, feedforward calculation of output of a trained neural network is very fast, making this approach particularly suitable for real-time applications. The network was trained on a large set of simulated equilibrium data for the DIII-D tokamak. The database encompasses a large variety of current profiles including the hollow current profiles important for reversed central shear operation. The parameters of safety factor profile (a quantity related to the current profile through the magnetic field tilt angle) estimated by the neural network include central safety factor, q0, minimum value of q, qmin, and the location of qmin. Very good performance of the trained neural network both for simulated test data and for experimental data is demonstrated.

Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cui, Z. Q.; Chen, Z. J.; Xie, X. F.

2014-11-15

The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic fieldmore » of 200 G.« less

Visco-Resistive MHD Modeling Benchmark of Forced Magnetic Reconnection

NASA Astrophysics Data System (ADS)

Beidler, M. T.; Hegna, C. C.; Sovinec, C. R.; Callen, J. D.; Ferraro, N. M.

2016-10-01

The presence of externally-applied 3D magnetic fields can affect important phenomena in tokamaks, including mode locking, disruptions, and edge localized modes. External fields penetrate into the plasma and can lead to forced magnetic reconnection (FMR), and hence magnetic islands, on resonant surfaces if the local plasma rotation relative to the external field is slow. Preliminary visco-resistive MHD simulations of FMR in a slab geometry are consistent with theory. Specifically, linear simulations exhibit proper scaling of the penetrated field with resistivity, viscosity, and flow, and nonlinear simulations exhibit a bifurcation from a flow-screened to a field-penetrated, magnetic island state as the external field is increased, due to the 3D electromagnetic force. These results will be compared to simulations of FMR in a circular cross-section, cylindrical geometry by way of a benchmark between the NIMROD and M3D-C1 extended-MHD codes. Because neither this geometry nor the MHD model has the physics of poloidal flow damping, the theory of will be expanded to include poloidal flow effects. The resulting theory will be tested with linear and nonlinear simulations that vary the resistivity, viscosity, flow, and external field. Supported by OFES DoE Grants DE-FG02-92ER54139, DE-FG02-86ER53218, DE-AC02-09CH11466, and the SciDAC Center for Extended MHD Modeling.

Edge-localized-modes in tokamaksa)

NASA Astrophysics Data System (ADS)

Leonard, A. W.

2014-09-01

Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively, rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heat flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. Encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.

Effects of finite poloidal gyroradius, shaping, and collisions on the zonal flow residuala)

NASA Astrophysics Data System (ADS)

Xiao, Yong; Catto, Peter J.; Dorland, William

2007-05-01

Zonal flow helps reduce and regulate the turbulent transport level in tokamaks. Rosenbluth and Hinton have shown that zonal flow damps to a nonvanishing residual level in collisionless [M. Rosenbluth and F. Hinton, Phys. Rev. Lett. 80, 724 (1998)] and collisional [F. Hinton and M. Rosenbluth, Plasma Phys. Control. Fusion 41, A653 (1999)] banana regime plasmas. Recent zonal flow advances are summarized including the evaluation of the effects on the zonal flow residual by plasma cross-section shaping, shorter wavelengths including those less than an electron gyroradius, and arbitrary ion collisionality relative to the zonal low frequency. In addition to giving a brief summary of these new developments, the analytic results are compared with GS2 numerical simulations [M. Kotschenreuther, G. Rewoldt, and W. Tang, Comput. Phys. Commun. 88, 128 (1991)] to demonstrate their value as benchmarks for turbulence codes.

Evaluating Stellarator Divertor Designs with EMC3

NASA Astrophysics Data System (ADS)

Bader, Aaron; Anderson, D. T.; Feng, Y.; Hegna, C. C.; Talmadge, J. N.

2013-10-01

In this paper various improvements of stellarator divertor design are explored. Next step stellarator devices require innovative divertor solutions to handle heat flux loads and impurity control. One avenue is to enhance magnetic flux expansion near strike points, somewhat akin to the X-Divertor concept in Tokamaks. The effect of judiciously placed external coils on flux deposition is calculated for configurations based on the HSX stellarator. In addition, we attempt to optimize divertor plate location to facilitate the external coil placement. Alternate areas of focus involve altering edge island size to elucidate the driving physics in the edge. The 3-D nature of stellarators complicates design and necessitates analysis of new divertor structures with appropriate simulation tools. We evaluate the various configurations with the coupled codes EMC3-EIRENE, allowing us to benchmark configurations based on target heat flux, impurity behavior, radiated power, and transitions to high recycling and detached regimes. Work supported by DOE-SC0006103.

The radial electric field dynamics in the neoclassical plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Novakovskii, S.V.; Liu, C.S.; Sagdeev, R.Z.

1997-12-01

A numerical simulation and analytical theory of the radial electric field dynamics in low collisional tokamak plasmas are presented. An initial value code {open_quotes}ELECTRIC{close_quotes} has been developed to solve the ion drift kinetic equation with a full collisional operator in the Hirshman{endash}Sigmar{endash}Clarke form together with the Maxwell equations. Different scenarios of relaxation of the radial electric field toward the steady-state in response to sudden and adiabatic changes of the equilibrium temperature gradient are presented. It is shown, that while the relaxation is usually accompanied by the geodesic acoustic oscillations, during the adiabatic change these oscillations are suppressed and only themore » magnetic pumping remains. Both the collisional damping and the Landau resonance interaction are shown to be important relaxation mechanisms. Scalings of the relaxation rates versus basic plasma parameters are presented. {copyright} {ital 1997 American Institute of Physics.}« less

Rigorous-two-Steps scheme of TRIPOLI-4® Monte Carlo code validation for shutdown dose rate calculation

NASA Astrophysics Data System (ADS)

Jaboulay, Jean-Charles; Brun, Emeric; Hugot, François-Xavier; Huynh, Tan-Dat; Malouch, Fadhel; Mancusi, Davide; Tsilanizara, Aime

2017-09-01

After fission or fusion reactor shutdown the activated structure emits decay photons. For maintenance operations the radiation dose map must be established in the reactor building. Several calculation schemes have been developed to calculate the shutdown dose rate. These schemes are widely developed in fusion application and more precisely for the ITER tokamak. This paper presents the rigorous-two-steps scheme implemented at CEA. It is based on the TRIPOLI-4® Monte Carlo code and the inventory code MENDEL. The ITER shutdown dose rate benchmark has been carried out, results are in a good agreement with the other participant.

The ARIES Advanced and Conservative Tokamak Power Plant Study

DOE PAGES

Kessel, C. E; Tillak, M. S; Najmabadi, F.; ...

2015-12-22

Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦ total N of 5.75, an H98 of 1.65,more » an n/n Gr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦ total N of 2.5, an H₉₈ of 1.25, an n/n Gr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.« less

The ARIES Advanced And Conservative Tokamak (ACT) Power Plant Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kessel, C. E.; Poli, F. M.; Ghantous, K.

2014-03-05

Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a βN total of 5.75, Hmore » 98 of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m 2. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βN total of 2.5, H 98 of 1.25, n/n Gr of 1.3, and peak divertor heat flux of 10 MW/m 2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m 2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.« less

Resistive Wall Modes Identification and Control in RFX-mod low qedge tokamak discharges

NASA Astrophysics Data System (ADS)

Baruzzo, Matteo; Bolzonella, Tommaso; Cavazzana, Roberto; Marchiori, Giuseppe; Marrelli, Lionello; Martin, Piero; Paccagnella, Roberto; Piovesan, Paolo; Piron, Lidia; Soppelsa, Anton; Zanca, Paolo; in, Yongkyoon; Liu, Yueqiang; Okabayashi, Michio; Takechi, Manabu; Villone, Fabio

2011-10-01

In this work the MHD stability of RFX mode tokamak discharges with qedge < 3 will be studied. The target plasma scenario is characterized by a plasma current 100kA

The effect of pressure anisotropy on ballooning modes in tokamak plasmas

NASA Astrophysics Data System (ADS)

Johnston, A.; Hole, M. J.; Qu, Z. S.; Hezaveh, H.

2018-06-01

Edge Localised Modes are thought to be caused by a spectrum of magnetohydrodynamic instabilities, including the ballooning mode. While ballooning modes have been studied extensively both theoretically and experimentally, the focus of the vast majority of this research has been on isotropic plasmas. The prevalence of pressure anisotropy in modern tokamaks thus motivates further study of these modes. This paper presents a numerical analysis of ballooning modes in anisotropic equilibria. The investigation was conducted using the newly-developed codes HELENA+ATF and MISHKA-A, which adds anisotropic physics to equilibria and stability analysis. We have examined the impact of anisotropy on the stability of an n = 30 ballooning mode, confirming results conform to previous calculations in the isotropic limit. Growth rates of ballooning modes in equilibria with different levels of anisotropy were then calculated using the stability code MISHKA-A. The key finding was that the level of anisotropy had a significant impact on ballooning mode growth rates. For {T}\\perp > {T}| | , typical of ICRH heating, the growth rate increases, while for {T}\\perp < {T}| | , typical of neutral beam heating, the growth rate decreases.

Experimental investigation of stability, frequency and toroidal mode number of compressional Alfvén eigenmodes in DIII-D

NASA Astrophysics Data System (ADS)

Tang, S.; Thome, K.; Pace, D.; Heidbrink, W. W.; Carter, T. A.; Crocker, N. A.; NSTX-U Collaboration; DIII-D Collaboration

2017-10-01

An experimental investigation of the stability of Doppler-shifted cyclotron resonant compressional Alfvén eigenmodes (CAE) using the flexible DIII-D neutral beams has begun to validate a theoretical understanding and realize the CAE's diagnostic potential. CAEs are excited by energetic ions from neutral beams [Heidbrink, NF 2006], with frequencies and toroidal mode numbers sensitive to the fast-ion phase space distribution, making them a potentially powerful passive diagnostic. The experiment also contributes to a predictive capability for spherical tokamak temperature profiles, where CAEs may play a role in energy transport [Crocker, NF 2013]. CAE activity was observed using the recently developed Ion Cyclotron Emission diagnostic-high bandwidth edge magnetic sensors sampled at 200 MS/s. Preliminary results show CAEs become unstable in BT ramp discharges below a critical threshold in the range 1.7 - 1.9 T, with the exact value increasing as density increases. The experiment will be used to validate simulations from relevant codes such as the Hybrid MHD code [Belova, PRL 2015]. This work was supported by US DOE Grants DE-SC0011810 and DE-FC02-04ER54698.

Effect of electron-to-ion mass ratio on radial electric field generation in tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Zhenqian; Dong, Jiaqi; Sheng, Zhengmao

Generation of coherent radial electric fields in plasma by drift-wave turbulence driven by plasma inhomogeneities is ab initio studied using gyro-kinetic particle simulation for conditions of operational tokamaks. In particular, the effect of the electron-to-ion mass ratio epsilon on the entire evolution of the plasma is considered. In conclusion, it is found that the electric field can be increased, and the turbulence-induced particle transport reduced, by making epsilon smaller, in agreement with many existing experimental observations.

Effect of electron-to-ion mass ratio on radial electric field generation in tokamak

DOE PAGES

Li, Zhenqian; Dong, Jiaqi; Sheng, Zhengmao; ...

2017-11-21

Generation of coherent radial electric fields in plasma by drift-wave turbulence driven by plasma inhomogeneities is ab initio studied using gyro-kinetic particle simulation for conditions of operational tokamaks. In particular, the effect of the electron-to-ion mass ratio epsilon on the entire evolution of the plasma is considered. In conclusion, it is found that the electric field can be increased, and the turbulence-induced particle transport reduced, by making epsilon smaller, in agreement with many existing experimental observations.

Beyond Orbital-Motion-Limited theory effects for dust transport in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Delzanno, Gian Luca; Tang, Xianzhu

Dust transport in tokamaks is very important for ITER. Can many kilograms of dust really accumulate in the device? Can the dust survive? The conventional dust transport model is based on Orbital-Motion-Limited theory (OML). But OML can break in the limit where the dust grain becomes positively charged due to electron emission processes because it overestimates the dust collected power. An OML + approximation of the emitted electrons trapped/passing boundary is shown to be in good agreement with PIC simulations.

Generation of noninductive current by electron-Bernstein waves on the COMPASS-D Tokamak.

PubMed

Shevchenko, V; Baranov, Y; O'Brien, M; Saveliev, A

2002-12-23

Electron-Bernstein waves (EBW) were excited in the plasma by mode converted extraordinary (X) waves launched from the high field side of the COMPASS-D tokamak at different toroidal angles. It has been found experimentally that X-mode injection perpendicular to the magnetic field provides maximum heating efficiency. Noninductive currents of up to 100 kA were found to be driven by the EBW mode with countercurrent drive. These results are consistent with ray tracing and quasilinear Fokker-Planck simulations.

Small angle slot divertor concept for long pulse advanced tokamaks

NASA Astrophysics Data System (ADS)

Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.

2017-04-01

SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.

The interaction between fishbone modes and shear Alfvén waves in tokamak plasmas

NASA Astrophysics Data System (ADS)

He, Hongda; Liu, Yueqiang; Dong, J. Q.; Hao, G. Z.; Wu, Tingting; He, Zhixiong; Zhao, K.

2016-05-01

The resonant interaction between the energetic particle triggered fishbone mode and the shear Alfvén waves is computationally investigated and firmly demonstrated based on a tokamak plasma equilibrium, using the self-consistent MHD-kinetic hybrid code MARS-K (Liu et al 2008 Phys. Plasmas 15 112503). This type of continuum resonance, occurring critically due to the mode’s toroidal rotation in the plasma frame, significantly modifies the eigenmode structure of the fishbone instability, by introducing two large peaks of the perturbed parallel current density near but offside the q  =  1 rational surface (q is the safety factor). The self-consistently computed radial plasma displacement substantially differs from that being assumed in the conventional fishbone theory.

A Finite-Orbit-Width Fokker-Planck solver for modeling of energetic particle interactions with waves, with application to Helicons in ITER

NASA Astrophysics Data System (ADS)

Petrov, Yuri V.; Harvey, R. W.

2017-10-01

The bounce-average (BA) finite-difference Fokker-Planck (FP) code CQL3D [1,2] now includes the essential physics to describe the RF heating of Finite-Orbit-Width (FOW) ions in tokamaks. The FP equation is reformulated in terms of Constants-Of-Motion coordinates, which we select to be particle speed, pitch angle, and major radius on the equatorial plane thus obtaining the distribution function directly at this location. Full-orbit, low collisionality neoclassical radial transport emerges from averaging the local friction and diffusion coefficients along guiding center orbits. Similarly, the BA of local quasilinear RF diffusion terms gives rise to additional radial transport. The local RF electric field components needed for the BA operator are usually obtained by a ray-tracing code, such as GENRAY, or in conjunction with full-wave codes. As a new, practical application, the CQL3D-FOW version is used for simulation of alpha-particle heating by high-harmonic waves in ITER. Coupling of high harmonic or helicon fast waves power to electrons is a promising current drive (CD) scenario for high beta plasmas. However, the efficiency of current drive can be diminished by parasitic channeling of RF power into fast ions, such as alphas, through finite Larmor-radius effects. We investigate possibilities to reduce the fast ion heating in CD scenarios.

Stray light analysis for the Thomson scattering diagnostic of the ETE Tokamak.

PubMed

Berni, L A; Albuquerque, B F C

2010-12-01

Thomson scattering is a well-established diagnostic for measuring local electron temperature and density in fusion plasma, but this technique is particularly difficult to implement due to stray light that can easily mask the scattered signal from plasma. To mitigate this problem in the multipoint Thomson scattering system implemented at the ETE (Experimento Tokamak Esférico) a detailed stray light analysis was performed. The diagnostic system was simulated in ZEMAX software and scattering profiles of the mechanical parts were measured in the laboratory in order to have near realistic results. From simulation, it was possible to identify the main points that contribute to the stray signals and changes in the dump were implemented reducing the stray light signals up to 60 times.

Fast low-to-high confinement mode bifurcation dynamics in a tokamak edge plasma gyrokinetic simulation

DOE PAGES

Chang, C. S.; Ku, S.; Tynan, G. R.; ...

2017-04-25

Transport barrier formation and its relation to sheared flows in fluids and plasmas are of fundamental interest in various natural and laboratory observations and of critical importance in achieving an economical energy production in a magnetic fusion device. Here we report the first observation of an edge transport barrier formation event in an electrostatic gyrokinetic simulation carried out in a realistic diverted tokamak edge geometry under strong forcing by a high rate of heat deposition. Here, the results show that turbulent Reynolds-stress-driven sheared E x B flows act in concert with neoclassical orbit loss to quench turbulent transport and formmore » a transport barrier just inside the last closed magnetic flux surface.« less

Velocity shear, turbulent saturation, and steep plasma gradients in the scrape-off layer of inner-wall limited tokamaks

DOE PAGES

Halpern, Federico D.; Ricci, Paolo

2016-12-19

The narrow power decay-length (λ q), recently found in the scrape-off layer (SOL) of inner wall limited (IWL) discharges in tokamaks, is studied using 3D, flux-driven, global two fluid turbulence simulations. The formation of the steep plasma profiles is found to arise due to radially sheared E×B poloidal flows. A complex interaction between sheared flows and parallel plasma currents outflowing into the sheath regulates the turbulent saturation, determining the transport levels. We quantify the effects of sheared flows, obtaining theoretical estimates in agreement with our non-linear simulations. As a result, analytical calculations suggest that the IWL λ q is roughlymore » equal to the turbulent correlation length.« less

Approach to ignition of tokamak reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sigmar, D.J.

1981-02-01

Recent transport modeling results for JET, INTOR, and ETF are reviewed and analyzed with respect to existing uncertainties in the underlying physics, the self-consistency of the very large numerical codes, and the margin for ignition. The codes show ignition to occur in ETF/INTOR-sized machines if empirical scaling can be extrapolated to ion temperatures (and beta values) much higher than those presently achieved, if there is no significant impurity accumulation over the first 7 s, and if the known ideal and resistive MHD instabilities remain controllable for the evolving plasma profiles during ignition startup.

Spheromak reactor-design study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Les, J.M.

1981-06-30

A general overview of spheromak reactor characteristics, such as MHD stability, start up, and plasma geometry is presented. In addition, comparisons are made between spheromaks, tokamaks and field reversed mirrors. The computer code Sphero is also discussed. Sphero is a zero dimensional time independent transport code that uses particle confinement times and profile parameters as input since they are not known with certainty at the present time. More specifically, Sphero numerically solves a given set of transport equations whose solutions include such variables as fuel ion (deuterium and tritium) density, electron density, alpha particle density and ion, electron temperatures.

Plasma simulations that meet the challenges of HST & JWST Active Nuclei & Starburst observations

NASA Astrophysics Data System (ADS)

Ferland, Gary

2017-08-01

Recent HST AGN monitoring programs, such as the STORM Campaign, have resulted in the definitive set of emission-line-continuum lag measurements. The goals are to measure the structure of the inner regions of an AGN, understand the physics driving the variability, and use this to place black hole mass determinations on an even firmer footing. Photoionization models make it possible to convert these observations into physical parameters such as cloud density or location. Here I propose to improve the treatment of emission from species like C IV, C III], Mg II, or Fe II in the spectral / plasma simulation code Cloudy. Like all plasma codes, Cloudy uses a modified two-level approximation to solve for the ionization of many-electron ions. I have participated in meetings on modeling Tokamak plasmas, which share many of the properties of the BLR of AGN and have the advantage of being a controlled laboratory environment. These discussions have led to the development of tests to show the density range over which the two-level approximation is valid. It fails at the densities where the strong UV lines form. I will use the atomic data available within the fusion modeling community, along with the methods they have developed, to improve Cloudy models so that they can better inform us of the message in the UV spectrum. The improvements will be part of future releases of Cloudy, which is openly available and updated on a regular basis.

Synthetic NPA diagnostic for energetic particles in JET plasmas

NASA Astrophysics Data System (ADS)

Varje, J.; Sirén, P.; Weisen, H.; Kurki-Suonio, T.; Äkäslompolo, S.; contributors, JET

2017-11-01

Neutral particle analysis (NPA) is one of the few methods for diagnosing fast ions inside a plasma by measuring neutral atom fluxes emitted due to charge exchange reactions. The JET tokamak features an NPA diagnostic which measures neutral atom fluxes and energy spectra simultaneously for hydrogen, deuterium and tritium species. A synthetic NPA diagnostic has been developed and used to interpret these measurements to diagnose energetic particles in JET plasmas with neutral beam injection (NBI) heating. The synthetic NPA diagnostic performs a Monte Carlo calculation of the neutral atom fluxes in a realistic geometry. The 4D fast ion distributions, representing NBI ions, were simulated using the Monte Carlo orbit-following code ASCOT. Neutral atom density profiles were calculated using the FRANTIC neutral code in the JINTRAC modelling suite. Additionally, for rapid analysis, a scan of neutral profiles was precalculated with FRANTIC for a range of typical plasma parameters. These were taken from the JETPEAK database, which includes a comprehensive set of data from the flat-top phases of nearly all discharges in recent JET campaigns. The synthetic diagnostic was applied to various JET plasmas in the recent hydrogen campaign where different hydrogen/deuterium mixtures and NBI configurations were used. The simulated neutral fluxes from the fast ion distributions were found to agree with the measured fluxes, reproducing the slowing-down profiles for different beam isotopes and energies and quantitatively estimating the fraction of hydrogen and deuterium fast ions.

Nonlinear Fluid Model Of 3-D Field Effects In Tokamak Plasmas

NASA Astrophysics Data System (ADS)

Callen, J. D.; Hegna, C. C.; Beidler, M. T.

2017-10-01

Extended MHD codes (e.g., NIMROD, M3D-C1) are beginning to explore nonlinear effects of small 3-D magnetic fields on tokamak plasmas. To facilitate development of analogous physically understandable reduced models, a fluid-based dynamic nonlinear model of these added 3-D field effects in the base axisymmetric tokamak magnetic field geometry is being developed. The model incorporates kinetic-based closures within an extended MHD framework. Key 3-D field effects models that have been developed include: 1) a comprehensive modified Rutherford equation for the growth of a magnetic island that includes the classical tearing and NTM perturbed bootstrap current drives, externally applied magnetic field and current drives, and classical and neoclassical polarization current effects, and 2) dynamic nonlinear evolution of the plasma toroidal flow (radial electric field) in response to the 3-D fields. An application of this model to RMP ELM suppression precipitated by an ELM crash will be discussed. Supported by Office of Fusion Energy Sciences, Office of Science, Dept. of Energy Grants DE-FG02-86ER53218 and DE-FG02-92ER54139.

MHD Studies of Advanced Tokamak Equilibria

NASA Astrophysics Data System (ADS)

Strumberger, E.

2005-10-01

Advanced tokamak scenarios are often characterized by an extremely reversed profile of the safety factor, q, and a fast toroidal rotation. ASDEX Upgrade type equilibria with toroidal flow are computed up to a toroidal Mach number of Mta= 0.5, and compared with the static solution. Using these equilibria, the stabilizing effect of differential toroidal rotation on double tearing modes (DTMs) is investigated. These studies show that the computation of equilibria with flow is necessary for toroidally rotating plasma with Mta>=0.2. The use of ρtor instead of ρpol as radial coordinate enables us also to investigate the stability of equilibria with current holes. For numerical reasons, the rotational transform, = 1/q, has to be unequal zero in the CASTOR$FLOW code, but values of a>=0.001 (qa<=1000) can be easily handled. Stability studies of DTMs in the presence of a current hole are presented. Tokamak equilibria are only approximately axisymmetric. The finite number of toroidal field coils destroys the perfect axisymmetry of the device, and the coils produce a short wavelength ripple in the magnetic field strength. This toroidal field ripple plays a crucial role for the loss of high energy particles. Therefore, three-dimensional tokamak equilibria with and without current holes are computed for various plasma beta values. In addition the influence of the plasma beta on the toroidal field ripple is investigated.

Integrated modeling of plasma ramp-up in DIII-D ITER-like and high bootstrap current scenario discharges

NASA Astrophysics Data System (ADS)

Wu, M. Q.; Pan, C. K.; Chan, V. S.; Li, G. Q.; Garofalo, A. M.; Jian, X.; Liu, L.; Ren, Q. L.; Chen, J. L.; Gao, X.; Gong, X. Z.; Ding, S. Y.; Qian, J. P.; Cfetr Physics Team

2018-04-01

Time-dependent integrated modeling of DIII-D ITER-like and high bootstrap current plasma ramp-up discharges has been performed with the equilibrium code EFIT, and the transport codes TGYRO and ONETWO. Electron and ion temperature profiles are simulated by TGYRO with the TGLF (SAT0 or VX model) turbulent and NEO neoclassical transport models. The VX model is a new empirical extension of the TGLF turbulent model [Jian et al., Nucl. Fusion 58, 016011 (2018)], which captures the physics of multi-scale interaction between low-k and high-k turbulence from nonlinear gyro-kinetic simulation. This model is demonstrated to accurately model low Ip discharges from the EAST tokamak. Time evolution of the plasma current density profile is simulated by ONETWO with the experimental current ramp-up rate. The general trend of the predicted evolution of the current density profile is consistent with that obtained from the equilibrium reconstruction with Motional Stark effect constraints. The predicted evolution of βN , li , and βP also agrees well with the experiments. For the ITER-like cases, the predicted electron and ion temperature profiles using TGLF_Sat0 agree closely with the experimental measured profiles, and are demonstrably better than other proposed transport models. For the high bootstrap current case, the predicted electron and ion temperature profiles perform better in the VX model. It is found that the SAT0 model works well at high IP (>0.76 MA) while the VX model covers a wider range of plasma current ( IP > 0.6 MA). The results reported in this paper suggest that the developed integrated modeling could be a candidate for ITER and CFETR ramp-up engineering design modeling.

Use of high order, periodic orbits in the PIES code

NASA Astrophysics Data System (ADS)

Monticello, Donald; Reiman, Allan

2010-11-01

We have implemented a version of the PIES code (Princeton Iterative Equilibrium SolverootnotetextA. Reiman et al 2007 Nucl. Fusion 47 572) that uses high order periodic orbits to select the surfaces on which straight magnetic field line coordinates will be calculated. The use of high order periodic orbits has increase the robustness and speed of the PIES code. We now have more uniform treatment of in-phase and out-of-phase islands. This new version has better convergence properties and works well with a full Newton scheme. We now have the ability to shrink islands using a bootstrap like current and this includes the m=1 island in tokamaks.

Hybrid simulation of fishbone instabilities in the EAST tokamak

NASA Astrophysics Data System (ADS)

Shen, Wei; Fu, Guoyong; Wang, Feng; Xu, Liqing; Li, Guoqiang; Liu, Chengyue; EAST Team

2017-10-01

Hybrid simulations with the global kinetic- MHD code M3D-K have been carried out to investigate the linear stability and nonlinear dynamics of beam-driven fishbone in EAST experiment. Linear simulations show that a low frequency fishbone instability is excited at experimental value of beam ion pressure. The mode is mainly driven by low energy beam ions via precessional resonance. The results are consistent with the experimental measurement with respect to mode frequency and mode structure. When the beam ion pressure is increased to exceed a critical value, the low frequency mode transits to a BAE with much higher frequency. Nonlinear simulations show that the frequency of the low frequency fishbone chirps up and down with corresponding hole-clump structures in phase space, consistent with the Berk-Breizman theory. In addition to the low frequency mode, the high frequency BAE is excited during the nonlinear evolution. For the transient case of beam pressure fraction where the low and high frequency modes are simultaneously excited in the linear phase, only one dominant mode appears in the nonlinear phase with frequency jumps up and down during nonlinear evolution. This work is supported by the National Natural Science Foundation of China under Grant Nos. 11605245 and 11505022, and the CASHIPS Director's Fund under Grant No. YZJJ201510, and the Department of Energy Scientific Discovery through Advanced Computing (SciDAC) under Grant No. DE-AC02-09CH11466.

Towards a 1 MW, 170 GHz gyrotron design for fusion application

NASA Astrophysics Data System (ADS)

Kumar, Anil; Kumar, Nitin; Singh, Udaybir; Bhattacharya, Ranajoy; Yadav, Vivek; Sinha, A. K.

2013-03-01

The electrical design of different components of 1 MW, 170 GHz gyrotron such as, magnetron injection gun, cylindrical interaction cavity and collector and RF window is presented in this article. Recently, a new project related to the development of 170 GHz, 1 MW gyrotron has been started for the Indian Tokamak. TE34,10 mode is selected as the operating mode after studied the problem of mode competition. The triode type geometry is selected for the design of magnetron injection gun (MIG) to achieve the required beam parameters. The maximum transverse velocity spread of 3.28% at the velocity ratio of 1.34 is obtained in simulations for a 40 A, 80 kV electron beam. The RF output power of more than 1 MW with 36.5% interaction efficiency without depressed collector is predicted by simulation in single-mode operation at 170 GHz frequency. The simulated single-stage depressed collector of the gyrotron predicted the overall device efficiencies >55%. Due to the very good thermal conductivity and very weak dependency of the dielectric parameters on temperature, PACVD diamond is selected for window design for the transmission of RF power. The in-house developed code MIGSYN and GCOMS are used for initial geometry design of MIG and mode selection respectively. Commercially available simulation tools MAGIC and ANSYS are used for beam-wave interaction and mechanical analysis respectively.

Tractable flux-driven temperature, density, and rotation profile evolution with the quasilinear gyrokinetic transport model QuaLiKiz

NASA Astrophysics Data System (ADS)

Citrin, J.; Bourdelle, C.; Casson, F. J.; Angioni, C.; Bonanomi, N.; Camenen, Y.; Garbet, X.; Garzotti, L.; Görler, T.; Gürcan, O.; Koechl, F.; Imbeaux, F.; Linder, O.; van de Plassche, K.; Strand, P.; Szepesi, G.; Contributors, JET

2017-12-01

Quasilinear turbulent transport models are a successful tool for prediction of core tokamak plasma profiles in many regimes. Their success hinges on the reproduction of local nonlinear gyrokinetic fluxes. We focus on significant progress in the quasilinear gyrokinetic transport model QuaLiKiz (Bourdelle et al 2016 Plasma Phys. Control. Fusion 58 014036), which employs an approximated solution of the mode structures to significantly speed up computation time compared to full linear gyrokinetic solvers. Optimisation of the dispersion relation solution algorithm within integrated modelling applications leads to flux calculations × {10}6-7 faster than local nonlinear simulations. This allows tractable simulation of flux-driven dynamic profile evolution including all transport channels: ion and electron heat, main particles, impurities, and momentum. Furthermore, QuaLiKiz now includes the impact of rotation and temperature anisotropy induced poloidal asymmetry on heavy impurity transport, important for W-transport applications. Application within the JETTO integrated modelling code results in 1 s of JET plasma simulation within 10 h using 10 CPUs. Simultaneous predictions of core density, temperature, and toroidal rotation profiles for both JET hybrid and baseline experiments are presented, covering both ion and electron turbulence scales. The simulations are successfully compared to measured profiles, with agreement mostly in the 5%-25% range according to standard figures of merit. QuaLiKiz is now open source and available at www.qualikiz.com.

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

DOE PAGES

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.; ...

2017-06-22

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

Stray light analysis for the Thomson scattering diagnostic of the ETE Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Berni, L. A.; Albuquerque, B. F. C.

2010-12-15

Thomson scattering is a well-established diagnostic for measuring local electron temperature and density in fusion plasma, but this technique is particularly difficult to implement due to stray light that can easily mask the scattered signal from plasma. To mitigate this problem in the multipoint Thomson scattering system implemented at the ETE (Experimento Tokamak Esferico) a detailed stray light analysis was performed. The diagnostic system was simulated in ZEMAX software and scattering profiles of the mechanical parts were measured in the laboratory in order to have near realistic results. From simulation, it was possible to identify the main points that contributemore » to the stray signals and changes in the dump were implemented reducing the stray light signals up to 60 times.« less

Two-fluid and magnetohydrodynamic modelling of magnetic reconnection in the MAST spherical tokamak and the solar corona

NASA Astrophysics Data System (ADS)

Browning, P. K.; Cardnell, S.; Evans, M.; Arese Lucini, F.; Lukin, V. S.; McClements, K. G.; Stanier, A.

2016-01-01

Twisted magnetic flux ropes are ubiquitous in laboratory and astrophysical plasmas, and the merging of such flux ropes through magnetic reconnection is an important mechanism for restructuring magnetic fields and releasing free magnetic energy. The merging-compression scenario is one possible start-up scheme for spherical tokamaks, which has been used on the Mega Amp Spherical Tokamak (MAST). Two current-carrying plasma rings or flux ropes approach each due to mutual attraction, forming a current sheet and subsequently merge through magnetic reconnection into a single plasma torus, with substantial plasma heating. Two-dimensional resistive and Hall-magnetohydrodynamic simulations of this process are reported, including a strong guide field. A model of the merging based on helicity-conserving relaxation to a minimum energy state is also presented, extending previous work to tight-aspect-ratio toroidal geometry. This model leads to a prediction of the final state of the merging, in good agreement with simulations and experiment, as well as the average temperature rise. A relaxation model of reconnection between two or more flux ropes in the solar corona is also described, allowing for different senses of twist, and the implications for heating of the solar corona are discussed.

Inductive flux usage and its optimization in tokamak operation

DOE PAGES

Luce, Timothy C.; Humphreys, David A.; Jackson, Gary L.; ...

2014-07-30

The energy flow from the poloidal field coils of a tokamak to the electromagnetic and kinetic stored energy of the plasma are considered in the context of optimizing the operation of ITER. The goal is to optimize the flux usage in order to allow the longest possible burn in ITER at the desired conditions to meet the physics objectives (500 MW fusion power with energy gain of 10). A mathematical formulation of the energy flow is derived and applied to experiments in the DIII-D tokamak that simulate the ITER design shape and relevant normalized current and pressure. The rate ofmore » rise of the plasma current was varied, and the fastest stable current rise is found to be the optimum for flux usage in DIII-D. A method to project the results to ITER is formulated. The constraints of the ITER poloidal field coil set yield an optimum at ramp rates slower than the maximum stable rate for plasmas similar to the DIII-D plasmas. Finally, experiments in present-day tokamaks for further optimization of the current rise and validation of the projections are suggested.« less

Three-dimensional magnetohydrodynamic equilibrium of quiescent H-modes in tokamak systems

NASA Astrophysics Data System (ADS)

Cooper, W. A.; Graves, J. P.; Duval, B. P.; Sauter, O.; Faustin, J. M.; Kleiner, A.; Lanthaler, S.; Patten, H.; Raghunathan, M.; Tran, T.-M.; Chapman, I. T.; Ham, C. J.

2016-06-01

Three dimensional free boundary magnetohydrodynamic equilibria that recover saturated ideal kink/peeling structures are obtained numerically. Simulations that model the JET tokamak at fixed < β > =1.7% with a large edge bootstrap current that flattens the q-profile near the plasma boundary demonstrate that a radial parallel current density ribbon with a dominant m /n  =  5/1 Fourier component at {{I}\\text{t}}=2.2 MA develops into a broadband spectrum when the toroidal current I t is increased to 2.5 MA.

Design of charge exchange recombination spectroscopy for the joint Texas experimental tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chi, Y.; Zhuang, G., E-mail: ge-zhuang@hust.edu.cn; Cheng, Z. F.

The old diagnostic neutral beam injector first operated at the University of Texas at Austin is ready for rejoining the joint Texas experimental tokamak (J-TEXT). A new set of high voltage power supplies has been equipped and there is no limitation for beam modulation or beam pulse duration henceforth. Based on the spectra of fully striped impurity ions induced by the diagnostic beam the design work for toroidal charge exchange recombination spectroscopy (CXRS) system is presented. The 529 nm carbon VI (n = 8 − 7 transition) line seems to be the best choice for ion temperature and plasma rotationmore » measurements and the considered hardware is listed. The design work of the toroidal CXRS system is guided by essential simulation of expected spectral results under the J-TEXT tokamak operation conditions.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ochs, I. E.; Bertelli, N.; Fisch, N. J.

Although lower hybrid waves are effective at driving currents in present-day tokamaks, they are expected to interact strongly with high-energy particles in extrapolating to reactors. In the presence of a radial alpha particle birth gradient, this interaction can take the form of wave amplification rather than damping. While it is known that this amplification more easily occurs when launching from the tokamak high-field side, the extent of this amplification has not been made quantitative. Here, by tracing rays launched from the high- field-side of a tokamak, the required radial gradients to achieve amplification are calculated for a temperature and densitymore » regime consistent with a hot-ion-mode fusion reactor. As a result, these simulations, while valid only in the linear regime of wave amplification, nonetheless illustrate the possibilities for wave amplification using high-field launch of the lower hybrid wave.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ochs, I. E.; Bertelli, N.; Fisch, N. J.

Although lower hybrid waves are effective at driving currents in present-day tokamaks, they are expected to interact strongly with high-energy particles in extrapolating to reactors. In the presence of a radial alpha particle birth gradient, this interaction can take the form of wave amplification rather than damping. While it is known that this amplification more easily occurs when launching from the tokamak high-field side, the extent of this amplification has not been made quantitative. Here, by tracing rays launched from the high-field-side of a tokamak, the required radial gradients to achieve amplification are calculated for a temperature and density regimemore » consistent with a hot-ion-mode fusion reactor. These simulations, while valid only in the linear regime of wave amplification, nonetheless illustrate the possibilities for wave amplification using high-field launch of the lower hybrid wave.« less

BESAFE II: Accident safety analysis code for MFE reactor designs

NASA Astrophysics Data System (ADS)

Sevigny, Lawrence Michael

The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications to BESAFE II is discussed in Chapter 6, for example, by adding additional environmental indices such as a waste disposal index. The biggest improvement to BESAFE II would be an increase in the database of activation product mobilization for a larger spectrum of fusion reactor materials. The ultimate goal we have is for BESAFE II to become part of a systems design program which would include economic factors and allow both safety and the cost of electricity to influence design.

Coupling of PIES 3-D Equilibrium Code and NIFS Bootstrap Code with Applications to the Computation of Stellarator Equilibria

NASA Astrophysics Data System (ADS)

Monticello, D. A.; Reiman, A. H.; Watanabe, K. Y.; Nakajima, N.; Okamoto, M.

1997-11-01

The existence of bootstrap currents in both tokamaks and stellarators was confirmed, experimentally, more than ten years ago. Such currents can have significant effects on the equilibrium and stability of these MHD devices. In addition, stellarators, with the notable exception of W7-X, are predicted to have such large bootstrap currents that reliable equilibrium calculations require the self-consistent evaluation of bootstrap currents. Modeling of discharges which contain islands requires an algorithm that does not assume good surfaces. Only one of the two 3-D equilibrium codes that exist, PIES( Reiman, A. H., Greenside, H. S., Compt. Phys. Commun. 43), (1986)., can easily be modified to handle bootstrap current. Here we report on the coupling of the PIES 3-D equilibrium code and NIFS bootstrap code(Watanabe, K., et al., Nuclear Fusion 35) (1995), 335.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schnack, D.D.; Lottati, I.; Mikic, Z.

The authors describe TRIM, a MHD code which uses finite volume discretization of the MHD equations on an unstructured adaptive grid of triangles in the poloidal plane. They apply it to problems related to modeling tokamak toroidal plasmas. The toroidal direction is treated by a pseudospectral method. Care was taken to center variables appropriately on the mesh and to construct a self adjoint diffusion operator for cell centered variables.

Simulations of thermionic suppression during tungsten transient melting experiments

NASA Astrophysics Data System (ADS)

Komm, M.; Tolias, P.; Ratynskaia, S.; Dejarnac, R.; Gunn, J. P.; Krieger, K.; Podolnik, A.; Pitts, R. A.; Panek, R.

2017-12-01

Plasma-facing components receive enormous heat fluxes under steady state and especially during transient conditions that can even lead to tungsten (W) melting. Under these conditions, the unimpeded thermionic current density emitted from the W surfaces can exceed the incident plasma current densities by several orders of magnitude triggering a replacement current which drives melt layer motion via the {\\boldsymbol{J}}× {\\boldsymbol{B}} force. However, in tokamaks, the thermionic current is suppressed by space-charge effects and prompt re-deposition due to gyro-rotation. We present comprehensive results of particle-in-cell modelling using the 2D3V code SPICE2 for the thermionic emissive sheath of tungsten. Simulations have been performed for various surface temperatures and selected inclinations of the magnetic field corresponding to the leading edge and sloped exposures. The surface temperature dependence of the escaping thermionic current and its limiting value are determined for various plasma parameters; for the leading edge geometry, the results agree remarkably well with the Takamura analytical model. For the sloped geometry, the limiting value is observed to be proportional to the thermal electron current and a simple analytical expression is proposed that accurately reproduces the numerical results.

Electron Scale Turbulence and Transport in an NSTX H-mode Plasma Using a Synthetic Diagnostic for High-k Scattering Measurements

NASA Astrophysics Data System (ADS)

Ruiz Ruiz, Juan; Guttenfelder, Walter; Loureiro, Nuno; Ren, Yang; White, Anne; MIT/PPPL Collaboration

2017-10-01

Turbulent fluctuations on the electron gyro-radius length scale are thought to cause anomalous transport of electron energy in spherical tokamaks such as NSTX and MAST in some parametric regimes. In NSTX, electron-scale turbulence is studied through a combination of experimental measurements from a high-k scattering system and gyrokinetic simulations. Until now most comparisons between experiment and simulation of electron scale turbulence have been qualitative, with recent work expanding to more quantitative comparisons via synthetic diagnostic development. In this new work, we propose two alternate, complementary ways to perform a synthetic diagnostic using the gyrokinetic code GYRO. The first approach builds on previous work and is based on the traditional selection of wavenumbers using a wavenumber filter, for which a new wavenumber mapping was implemented for general axisymmetric geometry. A second alternate approach selects wavenumbers in real-space to compute the power spectra. These approaches are complementary, and recent results from both synthetic diagnostic approaches applied to NSTX plasmas will be presented. Work supported by U.S. DOE contracts DE-AC02-09CH11466 and DE-AC02-05CH11231.

Automated divertor target design by adjoint shape sensitivity analysis and a one-shot method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dekeyser, W., E-mail: Wouter.Dekeyser@kuleuven.be; Reiter, D.; Baelmans, M.

As magnetic confinement fusion progresses towards the development of first reactor-scale devices, computational tokamak divertor design is a topic of high priority. Presently, edge plasma codes are used in a forward approach, where magnetic field and divertor geometry are manually adjusted to meet design requirements. Due to the complex edge plasma flows and large number of design variables, this method is computationally very demanding. On the other hand, efficient optimization-based design strategies have been developed in computational aerodynamics and fluid mechanics. Such an optimization approach to divertor target shape design is elaborated in the present paper. A general formulation ofmore » the design problems is given, and conditions characterizing the optimal designs are formulated. Using a continuous adjoint framework, design sensitivities can be computed at a cost of only two edge plasma simulations, independent of the number of design variables. Furthermore, by using a one-shot method the entire optimization problem can be solved at an equivalent cost of only a few forward simulations. The methodology is applied to target shape design for uniform power load, in simplified edge plasma geometry.« less

NSTX Disruption Simulations of Detailed Divertor and Passive Plate Models by Vector Potential Transfer from OPERA Global Analysis Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

P. H. Titus, S. Avasaralla, A.Brooks, R. Hatcher

2010-09-22

The National Spherical Torus Experiment (NSTX) project is planning upgrades to the toroidal field, plasma current and pulse length. This involves the replacement of the center-stack, including the inner legs of the TF, OH, and inner PF coils. A second neutral beam will also be added. The increased performance of the upgrade requires qualification of the remaining components including the vessel, passive plates, and divertor for higher disruption loads. The hardware needing qualification is more complex than is typically accessible by large scale electromagnetic (EM) simulations of the plasma disruptions. The usual method is to include simplified representations of componentsmore » in the large EM models and attempt to extract forces to apply to more detailed models. This paper describes a more efficient approach of combining comprehensive modeling of the plasma and tokamak conducting structures, using the 2D OPERA code, with much more detailed treatment of individual components using ANSYS electromagnetic (EM) and mechanical analysis. This capture local eddy currents and resulting loads in complex details, and allows efficient non-linear, and dynamic structural analyses.« less

Review of the 9th NLTE code comparison workshop

DOE PAGES

Piron, Robin; Gilleron, Franck; Aglitskiy, Yefim; ...

2017-02-24

Here, we review the 9th NLTE code comparison workshop, which was held in the Jussieu campus, Paris, from November 30th to December 4th, 2015. This time, the workshop was mainly focused on a systematic investigation of iron NLTE steady-state kinetics and emissivity, over a broad range of temperature and density. Through these comparisons, topics such as modeling of the dielectronic processes, density effects or the effect of an external radiation field were addressed. The K-shell spectroscopy of iron plasmas was also addressed, notably through the interpretation of tokamak and laser experimental spectra.

Review of the 9th NLTE code comparison workshop

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piron, Robin; Gilleron, Franck; Aglitskiy, Yefim

Here, we review the 9th NLTE code comparison workshop, which was held in the Jussieu campus, Paris, from November 30th to December 4th, 2015. This time, the workshop was mainly focused on a systematic investigation of iron NLTE steady-state kinetics and emissivity, over a broad range of temperature and density. Through these comparisons, topics such as modeling of the dielectronic processes, density effects or the effect of an external radiation field were addressed. The K-shell spectroscopy of iron plasmas was also addressed, notably through the interpretation of tokamak and laser experimental spectra.

Review of the 9th NLTE code comparison workshop

NASA Astrophysics Data System (ADS)

Piron, R.; Gilleron, F.; Aglitskiy, Y.; Chung, H.-K.; Fontes, C. J.; Hansen, S. B.; Marchuk, O.; Scott, H. A.; Stambulchik, E.; Ralchenko, Yu.

2017-06-01

We review the 9th NLTE code comparison workshop, which was held in the Jussieu campus, Paris, from November 30th to December 4th, 2015. This time, the workshop was mainly focused on a systematic investigation of iron NLTE steady-state kinetics and emissivity, over a broad range of temperature and density. Through these comparisons, topics such as modeling of the dielectronic processes, density effects or the effect of an external radiation field were addressed. The K-shell spectroscopy of iron plasmas was also addressed, notably through the interpretation of tokamak and laser experimental spectra.

Demonstration of Inductive Flux Saving by Transient CHI on NSTX

NASA Astrophysics Data System (ADS)

Raman, Roger

2010-11-01

Experiments in NSTX have now demonstrated the saving of central solenoid flux equivalent to 200kA of toroidal plasma current after coupling plasmas produced by Transient Coaxial Helicity Injection (CHI) to inductive sustainment and ramp-up of the toroidal plasma current [R. Raman, et al., PRL 104, 095003 (2010)]. This is a record for non-inductive plasma startup, and an important step for developing the spherical torus concept. With an injector current of only 4kA and total power supply energy of only 21 kJ, CHI initiated a toroidal current of 250 kA that when coupled to 0.11 Vs of induction ramped up to 525 kA without using any auxiliary heating, whereas an otherwise identical inductive-only discharge ramped to only 325 kA. This flux saving was realized by reducing the influx of low-Z impurities during the start-up phase through the use of electrode conditioning discharges, followed by lithium evaporative coating of the plasma-facing surfaces and reducing parasitic arcs in the upper divertor region through use of additional shaping-field coils. As a result of these improvements, and for the first time in NSTX, the electron temperature during the CHI phase continually increased with input energy, indicating that the additional injected energy was contributing to heating the plasma instead of being lost through impurity line radiation. Simulations with the Tokamak Simulation Code (TSC) show that the observed scaling of CHI start-up current with toroidal field in NSTX is consistent with theory, suggesting that use of CHI on larger machines is quite attractive. These exciting results from NSTX demonstrate that CHI is a viable solenoid-free plasma startup method for future STs and tokamaks. This work supported by U.S. DOE Contracts DE-AC02-09CH11466 and DE-FG02-99ER54519 AM08.

Dynamic simulations for preparing the acceptance test of JT-60SA cryogenic system

NASA Astrophysics Data System (ADS)

Cirillo, R.; Hoa, C.; Michel, F.; Poncet, J. M.; Rousset, B.

2016-12-01

Power generation in the future could be provided by thermo-nuclear fusion reactors like tokamaks. There inside, the fusion reaction takes place thanks to the generation of plasmas at hundreds of millions of degrees that must be confined magnetically with superconductive coils, cooled down to around 4.5 K. Within this frame, an experimental tokamak device, JT-60SA is currently under construction in Naka (Japan). The plasma works cyclically and the coil system is subject to pulsed heat loads. In order to size the refrigerator close to the average power and hence optimizing investment and operational costs, measures have to be taken to smooth the heat load. Here we present a dynamic model of the JT-60SA's Auxiliary Cold box (ACB) for preparing the acceptance tests of the refrigeration system planned in 2016 in Naka. The aim of this study is to simulate the pulsed load scenarios using different process controls. All the simulations have been performed with EcosimPro® and the associated cryogenic library: CRYOLIB.

OMFIT Tokamak Profile Data Fitting and Physics Analysis

DOE PAGES

Logan, N. C.; Grierson, B. A.; Haskey, S. R.; ...

2018-01-22

Here, One Modeling Framework for Integrated Tasks (OMFIT) has been used to develop a consistent tool for interfacing with, mapping, visualizing, and fitting tokamak profile measurements. OMFIT is used to integrate the many diverse diagnostics on multiple tokamak devices into a regular data structure, consistently applying spatial and temporal treatments to each channel of data. Tokamak data are fundamentally time dependent and are treated so from the start, with front-loaded and logic-based manipulations such as filtering based on the identification of edge-localized modes (ELMs) that commonly scatter data. Fitting is general in its approach, and tailorable in its application inmore » order to address physics constraints and handle the multiple spatial and temporal scales involved. Although community standard one-dimensional fitting is supported, including scale length–fitting and fitting polynomial-exponential blends to capture the H-mode pedestal, OMFITprofiles includes two-dimensional (2-D) fitting using bivariate splines or radial basis functions. These 2-D fits produce regular evolutions in time, removing jitter that has historically been smoothed ad hoc in transport applications. Profiles interface directly with a wide variety of models within the OMFIT framework, providing the inputs for TRANSP, kinetic-EFIT 2-D equilibrium, and GPEC three-dimensional equilibrium calculations. he OMFITprofiles tool’s rapid and comprehensive analysis of dynamic plasma profiles thus provides the critical link between raw tokamak data and simulations necessary for physics understanding.« less

OMFIT Tokamak Profile Data Fitting and Physics Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Logan, N. C.; Grierson, B. A.; Haskey, S. R.

Here, One Modeling Framework for Integrated Tasks (OMFIT) has been used to develop a consistent tool for interfacing with, mapping, visualizing, and fitting tokamak profile measurements. OMFIT is used to integrate the many diverse diagnostics on multiple tokamak devices into a regular data structure, consistently applying spatial and temporal treatments to each channel of data. Tokamak data are fundamentally time dependent and are treated so from the start, with front-loaded and logic-based manipulations such as filtering based on the identification of edge-localized modes (ELMs) that commonly scatter data. Fitting is general in its approach, and tailorable in its application inmore » order to address physics constraints and handle the multiple spatial and temporal scales involved. Although community standard one-dimensional fitting is supported, including scale length–fitting and fitting polynomial-exponential blends to capture the H-mode pedestal, OMFITprofiles includes two-dimensional (2-D) fitting using bivariate splines or radial basis functions. These 2-D fits produce regular evolutions in time, removing jitter that has historically been smoothed ad hoc in transport applications. Profiles interface directly with a wide variety of models within the OMFIT framework, providing the inputs for TRANSP, kinetic-EFIT 2-D equilibrium, and GPEC three-dimensional equilibrium calculations. he OMFITprofiles tool’s rapid and comprehensive analysis of dynamic plasma profiles thus provides the critical link between raw tokamak data and simulations necessary for physics understanding.« less

Dynamic optimization of open-loop input signals for ramp-up current profiles in tokamak plasmas

NASA Astrophysics Data System (ADS)

Ren, Zhigang; Xu, Chao; Lin, Qun; Loxton, Ryan; Teo, Kok Lay

2016-03-01

Establishing a good current spatial profile in tokamak fusion reactors is crucial to effective steady-state operation. The evolution of the current spatial profile is related to the evolution of the poloidal magnetic flux, which can be modeled in the normalized cylindrical coordinates using a parabolic partial differential equation (PDE) called the magnetic diffusion equation. In this paper, we consider the dynamic optimization problem of attaining the best possible current spatial profile during the ramp-up phase of the tokamak. We first use the Galerkin method to obtain a finite-dimensional ordinary differential equation (ODE) model based on the original magnetic diffusion PDE. Then, we combine the control parameterization method with a novel time-scaling transformation to obtain an approximate optimal parameter selection problem, which can be solved using gradient-based optimization techniques such as sequential quadratic programming (SQP). This control parameterization approach involves approximating the tokamak input signals by piecewise-linear functions whose slopes and break-points are decision variables to be optimized. We show that the gradient of the objective function with respect to the decision variables can be computed by solving an auxiliary dynamic system governing the state sensitivity matrix. Finally, we conclude the paper with simulation results for an example problem based on experimental data from the DIII-D tokamak in San Diego, California.

Statistical analysis and modeling of intermittent transport events in the tokamak scrape-off layer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anderson, Johan, E-mail: anderson.johan@gmail.com; Halpern, Federico D.; Ricci, Paolo

The turbulence observed in the scrape-off-layer of a tokamak is often characterized by intermittent events of bursty nature, a feature which raises concerns about the prediction of heat loads on the physical boundaries of the device. It appears thus necessary to delve into the statistical properties of turbulent physical fields such as density, electrostatic potential, and temperature, focusing on the mathematical expression of tails of the probability distribution functions. The method followed here is to generate statistical information from time-traces of the plasma density stemming from Braginskii-type fluid simulations and check this against a first-principles theoretical model. The analysis ofmore » the numerical simulations indicates that the probability distribution function of the intermittent process contains strong exponential tails, as predicted by the analytical theory.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vdovin V.L.

In this report we describe theory and 3D full wave code description for the wave excitation, propagation and absorption in 3-dimensional (3D) stellarator equilibrium high beta plasma in ion cyclotron frequency range (ICRF). This theory forms a basis for a 3D code creation, urgently needed for the ICRF heating scenarios development for the operated LHD, constructed W7-X, NCSX and projected CSX3 stellarators, as well for re evaluation of ICRF scenarios in operated tokamaks and in the ITER . The theory solves the 3D Maxwell-Vlasov antenna-plasma-conducting shell boundary value problem in the non-orthogonal flux coordinates ({Psi}, {theta}, {var_phi}), {Psi} being magneticmore » flux function, {theta} and {var_phi} being the poloidal and toroidal angles, respectively. All basic physics, like wave refraction, reflection and diffraction are self consistently included, along with the fundamental ion and ion minority cyclotron resonances, two ion hybrid resonance, electron Landau and TTMP absorption. Antenna reactive impedance and loading resistance are also calculated and urgently needed for an antenna -generator matching. This is accomplished in a real confining magnetic field being varying in a plasma major radius direction, in toroidal and poloidal directions, through making use of the hot dense plasma wave induced currents with account to the finite Larmor radius effects. We expand the solution in Fourier series over the toroidal ({var_phi}) and poloidal ({theta}) angles and solve resulting ordinary differential equations in a radial like {Psi}-coordinate by finite difference method. The constructed discretization scheme is divergent-free one, thus retaining the basic properties of original equations. The Fourier expansion over the angle coordinates has given to us the possibility to correctly construct the ''parallel'' wave number k{sub //}, and thereby to correctly describe the ICRF waves absorption by a hot plasma. The toroidal harmonics are tightly coupled with each other due to magnetic field inhomogeneity of stellarators in toroidal direction. This is drastically different from axial symmetric plasma of the tokamaks. The inclusion in the problem major radius variation of magnetic field can strongly modify earlier results obtained for the straight helical, especially for high beta plasma, due to location modification of the two ion hybrid resonance layers. For the NCSX, LHD, W7-AS and W7-X like magnetic field topology inclusion in our theory of a major radius inhomogeneity of the magnetic field is a key element for correct description of RF power deposition profiles at all. The theory is developed in a manner that includes tokamaks and magnetic mirrors as the particular cases through general metric tensor (provided by an equilibrium solver) treatment of the wave equations. We describe that newly developed stellarator ICRF 3D full wave code PSTELION, based on theory described in this report. Applications to tokamaks, ITER, stellarators and benchmarking with 2D TORIC and 3D AORSA codes are given in included subreports« less

Alpha channeling with high-field launch of lower hybrid waves

DOE PAGES

Ochs, I. E.; Bertelli, N.; Fisch, N. J.

2015-11-04

Although lower hybrid waves are effective at driving currents in present-day tokamaks, they are expected to interact strongly with high-energy particles in extrapolating to reactors. In the presence of a radial alpha particle birth gradient, this interaction can take the form of wave amplification rather than damping. While it is known that this amplification more easily occurs when launching from the tokamak high-field side, the extent of this amplification has not been made quantitative. Here, by tracing rays launched from the high- field-side of a tokamak, the required radial gradients to achieve amplification are calculated for a temperature and densitymore » regime consistent with a hot-ion-mode fusion reactor. As a result, these simulations, while valid only in the linear regime of wave amplification, nonetheless illustrate the possibilities for wave amplification using high-field launch of the lower hybrid wave.« less

Can tokamaks PFC survive a single event of any plasma instabilities?

NASA Astrophysics Data System (ADS)

Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

2013-07-01

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

Using AORSA to simulate helicon waves in DIIID and ITER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lau, Cornwall H; Jaeger, E. F.; Berry, Lee Alan

2014-01-01

Recent efforts by Vdovin [1] and Prater [2] have shown that helicon waves (fast waves at ~30 ion cyclotron frequency harmonic) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIIID, ITER and DEMO. For DIIID scenarios, the ray tracing code GENRAY has been extensively used to study helicon current drive efficiency and location as a function many plasma parameters. has some limitations on absorption at high cyclotron harmonics, so the full wave code AORSA, which is applicable to arbitrary Larmor radius and can therefore resolve high ion cyclotron harmonics, has been recentlymore » used to validate the GENRAY model. It will be shown that the GENRAY and AORSA driven current drive profiles are comparable for the envisioned high temperature and density advanced scenarios for DIIID, where there is high single pass absorption due to electron Landau damping. AORSA results will be shown for various plasma parameters for DIIID and for ITER. Computational difficulties in achieving these AORSA results will also be discussed. * Work supported by USDOE Contract No. DE-AC05-00OR22725 [1] V. L. Vdovin, Plasma Physics Reports, V.39, No.2, 2013 [2] R. Prater et al, Nucl. Fusion, 52, 083024, 2014« less

Ion temperature gradient driven transport in tokamaks with square shaping

DOE Office of Scientific and Technical Information (OSTI.GOV)

Joiner, N.; Dorland, W.

2010-06-15

Advanced tokamak schemes which may offer significant improvement to plasma confinement on the usual large aspect ratio Dee-shaped flux surface configuration are of great interest to the fusion community. One possibility is to introduce square shaping to the flux surfaces. The gyrokinetic code GS2[Kotschenreuther et al., Comput. Phys. Commun. 88, 128 (1996)] is used to study linear stability and the resulting nonlinear thermal transport of the ion temperature gradient driven (ITG) mode in tokamak equilibria with square shaping. The maximum linear growth rate of ITG modes is increased by negative squareness (diamond shaping) and reduced by positive values (square shaping).more » The dependence of thermal transport produced by saturated ITG instabilities on squareness is not as clear. The overall trend follows that of the linear instability, heat and particle fluxes increase with negative squareness and decrease with positive squareness. This is contradictory to recent experimental results [Holcomb et al., Phys. Plasmas 16, 056116 (2009)] which show a reduction in transport with negative squareness. This may be reconciled as a reduction in transport (consistent with the experiment) is observed at small negative values of the squareness parameter.« less

Equilibrium reconstruction with 3D eddy currents in the Lithium Tokamak eXperiment

DOE PAGES

Hansen, C.; Boyle, D. P.; Schmitt, J. C.; ...

2017-04-18

Axisymmetric free-boundary equilibrium reconstructions of tokamak plasmas in the Lithium Tokamak eXperiment (LTX) are performed using the PSI-Tri equilibrium code. Reconstructions in LTX are complicated by the presence of long-lived non-axisymmetric eddy currents generated by a vacuum vessel and first wall structures. To account for this effect, reconstructions are performed with additional toroidal current sources in these conducting regions. The eddy current sources are fixed in their poloidal distributions, but their magnitude is adjusted as part of the full reconstruction. Eddy distributions are computed by toroidally averaging currents, generated by coupling to vacuum field coils, from a simplified 3D filamentmore » model of important conducting structures. The full 3D eddy current fields are also used to enable the inclusion of local magnetic field measurements, which have strong 3D eddy current pick-up, as reconstruction constraints. Using this method, equilibrium reconstruction yields good agreement with all available diagnostic signals. Here, an accompanying field perturbation produced by 3D eddy currents on the plasma surface with a primarily n = 2, m = 1 character is also predicted for these equilibria.« less

A current drive by using the fast wave in frequency range higher than two timeslower hybrid resonance frequency on tokamaks

NASA Astrophysics Data System (ADS)

Kim, Sun Ho; Hwang, Yong Seok; Jeong, Seung Ho; Wang, Son Jong; Kwak, Jong Gu

2017-10-01

An efficient current drive scheme in central or off-axis region is required for the steady state operation of tokamak fusion reactors. The current drive by using the fast wave in frequency range higher than two times lower hybrid resonance (w>2wlh) could be such a scheme in high density, high temperature reactor-grade tokamak plasmas. First, it has relatively higher parallel electric field to the magnetic field favorable to the current generation, compared to fast waves in other frequency range. Second, it can deeply penetrate into high density plasmas compared to the slow wave in the same frequency range. Third, parasitic coupling to the slow wave can contribute also to the current drive avoiding parametric instability, thermal mode conversion and ion heating occured in the frequency range w<2wlh. In this study, the propagation boundary, accessibility, and the energy flow of the fast wave are given via cold dispersion relation and group velocity. The power absorption and current drive efficiency are discussed qualitatively through the hot dispersion relation and the polarization. Finally, those characteristics are confirmed with ray tracing code GENRAY for the KSTAR plasmas.

Transport and Dynamics in Toroidal Fusion Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sovinec, Carl

The study entitled, "Transport and Dynamics in Toroidal Fusion Systems," (TDTFS) applied analytical theory and numerical computation to investigate topics of importance to confining plasma, the fourth state of matter, with magnetic fields. A central focus of the work is how non-thermal components of the ion particle distribution affect the "sawtooth" collective oscillation in the core of the tokamak magnetic configuration. Previous experimental and analytical research had shown and described how the oscillation frequency decreases and amplitude increases, leading to "monster" or "giant" sawteeth, when the non-thermal component is increased by injecting particle beams or by exciting ions with imposedmore » electromagnetic waves. The TDTFS study applied numerical computation to self-consistently simulate the interaction between macroscopic collective plasma dynamics and the non-thermal particles. The modeling used the NIMROD code [Sovinec, Glasser, Gianakon, et al., J. Comput. Phys. 195, 355 (2004)] with the energetic component represented by simulation particles [Kim, Parker, Sovinec, and the NIMROD Team, Comput. Phys. Commun. 164, 448 (2004)]. The computations found decreasing growth rates for the instability that drives the oscillations, but they were ultimately limited from achieving experimentally relevant parameters due to computational practicalities. Nonetheless, this effort provided valuable lessons for integrated simulation of macroscopic plasma dynamics. It also motivated an investigation of the applicability of fluid-based modeling to the ion temperature gradient instability, leading to the journal publication [Schnack, Cheng, Barnes, and Parker, Phys. Plasmas 20, 062106 (2013)]. Apart from the tokamak-specific topics, the TDTFS study also addressed topics in the basic physics of magnetized plasma and in the dynamics of the reversed-field pinch (RFP) configuration. The basic physics work contributed to a study of two-fluid effects on interchange dynamics, where "two-fluid" refers to modeling independent dynamics of electron and ion species without full kinetic effects. In collaboration with scientist Ping Zhu, who received separate support, it was found that the rule-of-thumb criteria on stabilizing interchange has caveats that depend on the plasma density and temperature profiles. This work was published in [Zhu, Schnack, Ebrahimi, et al., Phys. Rev. Lett. 101, 085005 (2008)]. An investigation of general nonlinear relaxation with fluid models was partially supported by the TDTFS study and led to the publication [Khalzov, Ebrahimi, Schnack, and Mirnov, Phys. Plasmas 19, 012111 (2012)]. Work specific to the RFP included an investigation of interchange at large plasma pressure and support for applications [for example, Scheffel, Schnack, and Mirza, Nucl. Fusion 53, 113007 (2013)] of the DEBS code [Schnack, Barnes, Mikic, Harned, and Caramana, J. Comput. Phys. 70, 330 (1987)]. Finally, the principal investigator over most of the award period, Dalton Schnack, supervised a numerical study of modeling magnetic island suppression [Jenkins, Kruger, Hegna, Schnack, and Sovinec, Phys. Plasmas 17, 12502 (2010)].« less

Penetration of filamentary structures in the x-point region of spherical tokamaks

NASA Astrophysics Data System (ADS)

Baver, D. A.; Myra, J. R.; Scotti, F.; Zweben, S. J.; Militello, F.; Walkden, N.

2017-10-01

ArbiTER is a flexible eigenvalue code designed for plasma physics applications. It is used here to gain insight into the spatial dependence of filamentary structures in the scrape-off layer of spherical tokamaks. In particular, observations on MAST reveal the presence of a quiescent x-point region. Observations in NSTX similarly reveal a reduction in divertor fluctuations near the separatrix and a loss of midplane correlation. We will report on the penetration of filamentary structures into the vicinity of the x-point, as well as growth rate trends, for a variety of profiles and toroidal mode numbers. This will determine whether linear properties of these structures can explain experimental observations. Work supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-02ER54678.

2D electron density profile measurement in tokamak by laser-accelerated ion-beam probe.

PubMed

Chen, Y H; Yang, X Y; Lin, C; Wang, L; Xu, M; Wang, X G; Xiao, C J

2014-11-01

A new concept of Heavy Ion Beam Probe (HIBP) diagnostic has been proposed, of which the key is to replace the electrostatic accelerator of traditional HIBP by a laser-driven ion accelerator. Due to the large energy spread of ions, the laser-accelerated HIBP can measure the two-dimensional (2D) electron density profile of tokamak plasma. In a preliminary simulation, a 2D density profile was reconstructed with a spatial resolution of about 2 cm, and with the error below 15% in the core region. Diagnostics of 2D density fluctuation is also discussed.

DOE FES FY2017 Joint Research Target Fourth Quarter Milestone Report for theNational Spherical Torus Experiment Upgrade.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A.

2017-09-13

A successful high-performance plasma operation with a radiative divertor has been demonstrated on many tokamak devices, however, significant uncertainty remains in accurately modeling detachment thresholds, and in how detachment depends on divertor geometry. Whereas it was originally planned to perform dedicated divertor experiments on the National Spherical Tokamak Upgrade to address critical detachment and divertor geometry questions for this milestone, the experiments were deferred due to technical difficulties. Instead, existing NSTX divertor data was summarized and re-analyzed where applicable, and additional simulations were performed.

Modulation of Core Turbulent Density Fluctuations by Large-Scale Neoclassical Tearing Mode Islands in the DIII-D Tokamak

DOE PAGES

Bardóczi, L.; Rhodes, T. L.; Carter, T. A.; ...

2016-05-26

We report the first observation of localized modulation of turbulent density uctuations en (via Beam Emission Spectroscopy) by neoclassical tearing modes (NTMs) in the core of the DIII-D tokamak. NTMs are important as they often lead to severe degradation of plasma confinement and disruptions in high-confinement fusion experiments. Magnetic islands associated with NTMs significantly modify the profiles and turbulence drives. In this experiment n was found to be modulated by 14% across the island. Gyrokinetic simulations suggest that en could be dominantly driven by the ion temperature gradient (ITG) instability.

UCSD Performance in the Edge Plasma Simulation (EPSI) Project. Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tynan, George Robert

This report contains a final report on the activities of UC San Diego PI G.R. Tynan and his collaborators as part of the EPSI Project, that was led by Dr. C.S. Chang, from PPPL. As a part of our work, we carried out several experiments on the ALCATOR C-­MOD tokamak device, aimed at unraveling the “trigger” or cause of the spontaneous transition from low-­mode confinement (L-­mode) to high confinement (H-­mode) that is universally observed in tokamak devices, and is planned for use in ITER.

Magnetohydrodynamic modes analysis and control of Fusion Advanced Studies Torus high-current scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Villone, F.; Mastrostefano, S.; Calabrò, G.

2014-08-15

One of the main FAST (Fusion Advanced Studies Torus) goals is to have a flexible experiment capable to test tools and scenarios for safe and reliable tokamak operation, in order to support ITER and help the final DEMO design. In particular, in this paper, we focus on operation close to a possible border of stability related to low-q operation. To this purpose, a new FAST scenario has then been designed at I{sub p} = 10 MA, B{sub T} = 8.5 T, q{sub 95} ≈ 2.3. Transport simulations, carried out by using the code JETTO and the first principle transport model GLF23, indicate that, under these conditions, FASTmore » could achieve an equivalent Q ≈ 3.5. FAST will be equipped with a set of internal active coils for feedback control, which will produce magnetic perturbation with toroidal number n = 1 or n = 2. Magnetohydrodynamic (MHD) mode analysis and feedback control simulations performed with the codes MARS, MARS-F, CarMa (both assuming the presence of a perfect conductive wall and using the exact 3D resistive wall structure) show the possibility of the FAST conductive structures to stabilize n = 1 ideal modes. This leaves therefore room for active mitigation of the resistive mode (down to a characteristic time of 1 ms) for safety purposes, i.e., to avoid dangerous MHD-driven plasma disruption, when working close to the machine limits and magnetic and kinetic energy density not far from reactor values.« less

Development of ITER non-activation phase operation scenarios

DOE PAGES

Kim, S. H.; Poli, F. M.; Koechl, F.; ...

2017-06-29

Non-activation phase operations in ITER in hydrogen (H) and helium (He) will be important for commissioning of tokamak systems, such as diagnostics, heating and current drive (HCD) systems, coils and plasma control systems, and for validation of techniques necessary for establishing operations in DT. The assessment of feasible HCD schemes at various toroidal fields (2.65–5.3 T) has revealed that the previously applied assumptions need to be refined for the ITER non-activation phase H/He operations. A study of the ranges of plasma density and profile shape using the JINTRAC suite of codes has indicated that the hydrogen pellet fuelling into Hemore » plasmas should be utilized taking the optimization of IC power absorption, neutral beam shine-through density limit and H-mode access into account. The EPED1 estimation of the edge pedestal parameters has been extended to various H operation conditions, and the combined EPED1 and SOLPS estimation has provided guidance for modelling the edge pedestal in H/He operations. The availability of ITER HCD schemes, ranges of achievable plasma density and profile shape, and estimation of the edge pedestal parameters for H/He plasmas have been integrated into various time-dependent tokamak discharge simulations. In this paper, various H/He scenarios at a wide range of plasma current (7.5–15 MA) and field (2.65–5.3 T) have been developed for the ITER non-activation phase operation, and the sensitivity of the developed scenarios to the used assumptions has been investigated to provide guidance for further development.« less

Development of ITER non-activation phase operation scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, S. H.; Poli, F. M.; Koechl, F.

Non-activation phase operations in ITER in hydrogen (H) and helium (He) will be important for commissioning of tokamak systems, such as diagnostics, heating and current drive (HCD) systems, coils and plasma control systems, and for validation of techniques necessary for establishing operations in DT. The assessment of feasible HCD schemes at various toroidal fields (2.65–5.3 T) has revealed that the previously applied assumptions need to be refined for the ITER non-activation phase H/He operations. A study of the ranges of plasma density and profile shape using the JINTRAC suite of codes has indicated that the hydrogen pellet fuelling into Hemore » plasmas should be utilized taking the optimization of IC power absorption, neutral beam shine-through density limit and H-mode access into account. The EPED1 estimation of the edge pedestal parameters has been extended to various H operation conditions, and the combined EPED1 and SOLPS estimation has provided guidance for modelling the edge pedestal in H/He operations. The availability of ITER HCD schemes, ranges of achievable plasma density and profile shape, and estimation of the edge pedestal parameters for H/He plasmas have been integrated into various time-dependent tokamak discharge simulations. In this paper, various H/He scenarios at a wide range of plasma current (7.5–15 MA) and field (2.65–5.3 T) have been developed for the ITER non-activation phase operation, and the sensitivity of the developed scenarios to the used assumptions has been investigated to provide guidance for further development.« less

Quasilinear diffusion coefficients in a finite Larmor radius expansion for ion cyclotron heated plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Jungpyo; Wright, John; Bertelli, Nicola

In this study, a reduced model of quasilinear velocity diffusion by a small Larmor radius approximation is derived to couple the Maxwell’s equations and the Fokker Planck equation self-consistently for the ion cyclotron range of frequency waves in a tokamak. The reduced model ensures the important properties of the full model by Kennel-Engelmann diffusion, such as diffusion directions, wave polarizations, and H-theorem. The kinetic energy change (Wdot ) is used to derive the reduced model diffusion coefficients for the fundamental damping (n = 1) and the second harmonic damping (n = 2) to the lowest order of the finite Larmormore » radius expansion. The quasilinear diffusion coefficients are implemented in a coupled code (TORIC-CQL3D) with the equivalent reduced model of the dielectric tensor. We also present the simulations of the ITER minority heating scenario, in which the reduced model is verified within the allowable errors from the full model results.« less

Modeling of material erosion and redeposition for dedicated DiMES experiments on DIII-D

NASA Astrophysics Data System (ADS)

Ding, R.; Abrams, T.; Chrobak, C. P.; Guo, H. Y.; Snyder, P. B.; Chan, V. S.; Rudakov, D. L.; Stangeby, P. C.; Elder, J. D.; Tskhakaya, D.; Wampler, W. R.; Kirschner, A.; McLean, A. G.

2015-11-01

Erosion and redeposition of plasma facing materials is a key issue for high-power, long pulse tokamak operation. A series of experiments has been carried out on DIII-D in which well-characterized samples of different materials were exposed to divertor plasma using DiMES. Such experiments provide a good benchmark for PMI codes, such as ERO. It was found that the erosion and redeposition are strongly determined by the impurity content in the plasma and sheath properties near the surface. The principal experimental results (net erosion rate and profile, net/gross erosion ratio) are reproduced by ERO simulations to within the uncertainties, indicating that the controlling physics has likely been identified. New techniques suggested by modeling such as external biasing and local gas injection for suppressing material erosion are planned to be tested in DiMES/DIII-D experiments. Work supported by US DOE DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, DE-AC52-07NA27344.

Fishbone Mode Excited by Deeply Trapped Energetic Beam Ions in EAST

NASA Astrophysics Data System (ADS)

Zheng, Ting; Wu, Bin; Xu, Liqing; Hu, Chundong; Zang, Qing; Ding, Siye; Li, Yingying; Wu, Xingquan; Wang, Jinfang; Shen, Biao; Zhong, Guoqiang; Li, Hao; Shi, Tonghui; EAST Team

2016-06-01

This paper describes the fishbone mode phenomena during the injection of high-power neutral beams in EAST (Experimental Advanced Superconducting Tokamak). The features of the fishbone mode are presented. The change in frequency of the mode during a fishbone burst is from 1 kHz to 6 kHz. The nonlinear behavior of the fishbone mode is analyzed by using a prey-predator model, which is consistent with the experimental results. This model indicates that the periodic oscillations of the fishbone mode always occur near the critical value of fast ion beta. Furthermore, the neutral beam analysis for the discharge is done by using the NUBEAM module of the TRANSP code. According to the numerical simulation results and theoretical calculation, it can be concluded that the fishbone mode is driven by the deeply trapped energetic beam ions in EAST. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB101001, 2014DFG61950 and 2013GB112003) and National Natural Science Foundation of China (Nos. 11175211 and 11275233)

Quasilinear diffusion coefficients in a finite Larmor radius expansion for ion cyclotron heated plasmas

DOE PAGES

Lee, Jungpyo; Wright, John; Bertelli, Nicola; ...

2017-04-24

In this study, a reduced model of quasilinear velocity diffusion by a small Larmor radius approximation is derived to couple the Maxwell’s equations and the Fokker Planck equation self-consistently for the ion cyclotron range of frequency waves in a tokamak. The reduced model ensures the important properties of the full model by Kennel-Engelmann diffusion, such as diffusion directions, wave polarizations, and H-theorem. The kinetic energy change (Wdot ) is used to derive the reduced model diffusion coefficients for the fundamental damping (n = 1) and the second harmonic damping (n = 2) to the lowest order of the finite Larmormore » radius expansion. The quasilinear diffusion coefficients are implemented in a coupled code (TORIC-CQL3D) with the equivalent reduced model of the dielectric tensor. We also present the simulations of the ITER minority heating scenario, in which the reduced model is verified within the allowable errors from the full model results.« less

User's guide for GSMP, a General System Modeling Program. [In PL/I

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cook, J. M.

1979-10-01

GSMP is designed for use by systems analysis teams. Given compiled subroutines that model the behavior of components plus instructions as to how they are to be interconnected, this program links them together to model a complete system. GSMP offers a fast response to management requests for reconfigurations of old systems and even initial configurations of new systems. Standard system-analytic services are provided: parameter sweeps, graphics, free-form input and formatted output, file storage and recovery, user-tested error diagnostics, component model and integration checkout and debugging facilities, sensitivity analysis, and a multimethod optimizer with nonlinear constraint handling capability. Steady-state or cyclicmore » time-dependence is simulated directly, initial-value problems only indirectly. The code is written in PL/I, but interfaces well with FORTRAN component models. Over the last five years GSMP has been used to model theta-pinch, tokamak, and heavy-ion fusion power plants, open- and closed-cycle magneto-hydrodynamic power plants, and total community energy systems.« less

Computations of Vertical Displacement Events with Toroidal Asymmetry

NASA Astrophysics Data System (ADS)

Sovinec, C. R.; Bunkers, K. J.

2017-10-01

Nonlinear numerical MHD modeling with the NIMROD code [https://nimrodteam.org] is being developed to investigate asymmetry during vertical displacement events. We start from idealized up/down symmetric tokamak equilibria with small levels of imposed toroidally asymmetric field errors. Vertical displacement results when removing current from one of the two divertor coils. The Eulerian reference-frame modeling uses temperature-dependent resistivity and anisotropic thermal conduction to distinguish the hot plasma region from surrounding cold, low-density conditions. Diffusion through a resistive wall is slow relative to Alfvenic scales but much faster than resistive plasma diffusion. Loss of the initial edge pressure and current distributions leads to a narrow layer of parallel current, which drives low-n modes that may be related to peeling-dominated ELMs. These modes induce toroidal asymmetry in the conduction current, which connects the simulated plasma to the wall. Work supported by the US DOE through Grant Numbers DE-FG02-06ER54850 and DE-FC02-08ER54975.

The external kink mode in diverted tokamaks

NASA Astrophysics Data System (ADS)

Turnbull, A. D.; Hanson, J. M.; Turco, F.; Ferraro, N. M.; Lanctot, M. J.; Lao, L. L.; Strait, E. J.; Piovesan, P.; Martin, P.

2016-06-01

> . The resistive kink behaves much like the ideal kink with predominantly kink or interchange parity and no real sign of a tearing component. However, the growth rates scale with a fractional power of the resistivity near the surface. The results have a direct bearing on the conventional edge cutoff procedures used in most ideal MHD codes, as well as implications for ITER and for future reactor options.

Electromagnetic Torque in Tokamaks with Toroidal Asymmetries

NASA Astrophysics Data System (ADS)

Logan, Nikolas Christopher

Toroidal rotation and rotation shear strongly influences stability and confinement in tokamaks. Breaking of the toroidal symmetry by fields orders of magnitude smaller than the axisymmetric field can, however, produce electromagnetic torques that significantly affect the plasma rotation, stability and confinement. These electromagnetic torques are the study of this thesis. There are two typical types of electromagnetic torques in tokamaks: 1) "resonant torques" for which a plasma current defined by a single toroidal and single poloidal harmonic interact with external currents and 2) "nonresonant torques" for which the global plasma response to nonaxisymmetric fields is phase shifted by kinetic effects that drive the rotation towards a neoclassical offset. This work describes the diagnostics and analysis necessary to evaluate the torque by measuring the rate of momentum transfer per unit area in the vacuum region between the plasma and external currents using localized magnetic sensors to measure the Maxwell stress. These measurements provide model independent quantification of both the resonant and nonresonant electromagnetic torques, enabling direct verification of theoretical models. Measured values of the nonresonant torque are shown to agree well with the perturbed equilibrium nonambipolar transport (PENT) code calculation of torque from cross field transport in nonaxisymmetric equilibria. A combined neoclassical toroidal viscosity (NTV) theory, valid across a wide range of kinetic regimes, is fully implemented for the first time in general aspect ratio and shaped plasmas. The code captures pitch angle resonances, reproducing previously inaccessible collisionality limits in the model. The complete treatment of the model enables benchmarking to the hybrid kinetic MHD stability codes MARS-K and MISK, confirming the energy-torque equivalency principle in perturbed equilibria. Experimental validations of PENT results confirm the torque applied by nonaxisymmetric coils is often proportional to the energy put into the dominant ideal MHD kink mode. This reduces the control of nonresonant torque to a single mode model, enabling efficient feed forward optimization of applied fields. Initial results including the anisotropic kinetic pressure tensor directly in the plasma eigenmode calculations are presented here, and may eventually provide accurate metrics for multimodal coupling similar to the established single mode metrics.

Theoretical and experimental studies of a planar inductive coupled rf plasma source as the driver in simulator facility (ISTAPHM) of interactions of waves with the edge plasma on tokamaks

NASA Astrophysics Data System (ADS)

Ghanei, V.; Nasrabadi, M. N.; Chin, O.-H.; Jayapalan, K. K.

2017-11-01

This research aims to design and build a planar inductive coupled RF plasma source device which is the driver of the simulator project (ISTAPHM) of the interactions between ICRF Antenna and Plasma on tokamak by using the AMPICP model. For this purpose, a theoretical derivation of the distribution of the RF magnetic field in the plasma-filled reactor chamber is presented. An experimental investigation of the field distributions is described and Langmuir measurements are developed numerically. A comparison of theory and experiment provides an evaluation of plasma parameters in the planar ICP reactor. The objective of this study is to characterize the plasma produced by the source alone. We present the results of the first analysis of the plasma characteristics (plasma density, electron temperature, electron-ion collision frequency, particle fluxes and their velocities, stochastic frequency, skin depth and electron energy distribution functions) as function of the operating parameters (injected power, neutral pressure and magnetic field) as measured with fixed and movable Langmuir probes. The plasma is currently produced only by the planar ICP. The exact goal of these experiments is that the produced plasma by external source can exist as a plasma representative of the edge of tokamaks.