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.

Simulation of EAST vertical displacement events by tokamak simulation code

NASA Astrophysics Data System (ADS)

Qiu, Qinglai; Xiao, Bingjia; Guo, Yong; Liu, Lei; Xing, Zhe; Humphreys, D. A.

2016-10-01

Vertical instability is a potentially serious hazard for elongated plasma. In this paper, the tokamak simulation code (TSC) is used to simulate vertical displacement events (VDE) on the experimental advanced superconducting tokamak (EAST). Key parameters from simulations, including plasma current, plasma shape and position, flux contours and magnetic measurements match experimental data well. The growth rates simulated by TSC are in good agreement with TokSys results. In addition to modeling the free drift, an EAST fast vertical control model enables TSC to simulate the course of VDE recovery. The trajectories of the plasma current center and control currents on internal coils (IC) fit experimental data well.

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.

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.

The GBS code for tokamak scrape-off layer simulations

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

Halpern, F.D., E-mail: federico.halpern@epfl.ch; Ricci, P.; Jolliet, S.

2016-06-15

We describe a new version of GBS, a 3D global, flux-driven plasma turbulence code to simulate the turbulent dynamics in the tokamak scrape-off layer (SOL), superseding the code presented by Ricci et al. (2012) [14]. The present work is driven by the objective of studying SOL turbulent dynamics in medium size tokamaks and beyond with a high-fidelity physics model. We emphasize an intertwining framework of improved physics models and the computational improvements that allow them. The model extensions include neutral atom physics, finite ion temperature, the addition of a closed field line region, and a non-Boussinesq treatment of the polarizationmore » drift. GBS has been completely refactored with the introduction of a 3-D Cartesian communicator and a scalable parallel multigrid solver. We report dramatically enhanced parallel scalability, with the possibility of treating electromagnetic fluctuations very efficiently. The method of manufactured solutions as a verification process has been carried out for this new code version, demonstrating the correct implementation of the physical model.« less

Development of Numerical Methods to Estimate the Ohmic Breakdown Scenarios of a Tokamak

NASA Astrophysics Data System (ADS)

Yoo, Min-Gu; Kim, Jayhyun; An, Younghwa; Hwang, Yong-Seok; Shim, Seung Bo; Lee, Hae June; Na, Yong-Su

2011-10-01

The ohmic breakdown is a fundamental method to initiate the plasma in a tokamak. For the robust breakdown, ohmic breakdown scenarios have to be carefully designed by optimizing the magnetic field configurations to minimize the stray magnetic fields. This research focuses on development of numerical methods to estimate the ohmic breakdown scenarios by precise analysis of the magnetic field configurations. This is essential for the robust and optimal breakdown and start-up of fusion devices especially for ITER and its beyond equipped with low toroidal electric field (ET <= 0.3 V/m). A field-line-following analysis code based on the Townsend avalanche theory and a particle simulation code are developed to analyze the breakdown characteristics of actual complex magnetic field configurations including the stray magnetic fields in tokamaks. They are applied to the ohmic breakdown scenarios of tokamaks such as KSTAR and VEST and compared with experiments.

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.

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.

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.

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.

Advances in stellarator gyrokinetics

NASA Astrophysics Data System (ADS)

Helander, P.; Bird, T.; Jenko, F.; Kleiber, R.; Plunk, G. G.; Proll, J. H. E.; Riemann, J.; Xanthopoulos, P.

2015-05-01

Recent progress in the gyrokinetic theory of stellarator microinstabilities and turbulence simulations is summarized. The simulations have been carried out using two different gyrokinetic codes, the global particle-in-cell code EUTERPE and the continuum code GENE, which operates in the geometry of a flux tube or a flux surface but is local in the radial direction. Ion-temperature-gradient (ITG) and trapped-electron modes are studied and compared with their counterparts in axisymmetric tokamak geometry. Several interesting differences emerge. Because of the more complicated structure of the magnetic field, the fluctuations are much less evenly distributed over each flux surface in stellarators than in tokamaks. Instead of covering the entire outboard side of the torus, ITG turbulence is localized to narrow bands along the magnetic field in regions of unfavourable curvature, and the resulting transport depends on the normalized gyroradius ρ* even in radially local simulations. Trapped-electron modes can be significantly more stable than in typical tokamaks, because of the spatial separation of regions with trapped particles from those with bad magnetic curvature. Preliminary non-linear simulations in flux-tube geometry suggest differences in the turbulence levels in Wendelstein 7-X and a typical tokamak.

Three-dimensional analysis of tokamaks and stellarators

PubMed Central

Garabedian, Paul R.

2008-01-01

The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

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.

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

Hybrid parallelization of the XTOR-2F code for the simulation of two-fluid MHD instabilities in tokamaks

NASA Astrophysics Data System (ADS)

Marx, Alain; Lütjens, Hinrich

2017-03-01

A hybrid MPI/OpenMP parallel version of the XTOR-2F code [Lütjens and Luciani, J. Comput. Phys. 229 (2010) 8130] solving the two-fluid MHD equations in full tokamak geometry by means of an iterative Newton-Krylov matrix-free method has been developed. The present work shows that the code has been parallelized significantly despite the numerical profile of the problem solved by XTOR-2F, i.e. a discretization with pseudo-spectral representations in all angular directions, the stiffness of the two-fluid stability problem in tokamaks, and the use of a direct LU decomposition to invert the physical pre-conditioner at every Krylov iteration of the solver. The execution time of the parallelized version is an order of magnitude smaller than the sequential one for low resolution cases, with an increasing speedup when the discretization mesh is refined. Moreover, it allows to perform simulations with higher resolutions, previously forbidden because of memory limitations.

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

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.

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.

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.

Final Report Advanced Quasioptical Launcher System

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

Jeffrey Neilson

2010-04-30

This program developed an analytical design tool for designing antenna and mirror systems to convert whispering gallery RF modes to Gaussian or HE11 modes. Whispering gallery modes are generated by gyrotrons used for electron cyclotron heating of fusion plasmas in tokamaks. These modes cannot be easily transmitted and must be converted to free space or waveguide modes compatible with transmission line systems.This program improved the capability of SURF3D/LOT, which was initially developed in a previous SBIR program. This suite of codes revolutionized quasi-optical launcher design, and this code, or equivalent codes, are now used worldwide. This program added functionality tomore » SURF3D/LOT to allow creating of more compact launcher and mirror systems and provide direct coupling to corrugated waveguide within the vacuum envelope of the gyrotron. Analysis was also extended to include full-wave analysis of mirror transmission line systems. The code includes a graphical user interface and is available for advanced design of launcher systems.« less

Neutral helium beam probe

NASA Astrophysics Data System (ADS)

Karim, Rezwanul

1999-10-01

This article discusses the development of a code where diagnostic neutral helium beam can be used as a probe. The code solves numerically the evolution of the population densities of helium atoms at their several different energy levels as the beam propagates through the plasma. The collisional radiative model has been utilized in this numerical calculation. The spatial dependence of the metastable states of neutral helium atom, as obtained in this numerical analysis, offers a possible diagnostic tool for tokamak plasma. The spatial evolution for several hypothetical plasma conditions was tested. Simulation routines were also run with the plasma parameters (density and temperature profiles) similar to a shot in the Princeton beta experiment modified (PBX-M) tokamak and a shot in Tokamak Fusion Test Reactor tokamak. A comparison between the simulation result and the experimentally obtained data (for each of these two shots) is presented. A good correlation in such comparisons for a number of such shots can establish the accurateness and usefulness of this probe. The result can possibly be extended for other plasma machines and for various plasma conditions in those machines.

Multi-frequency ICRF diagnostic of Tokamak plasmas

NASA Astrophysics Data System (ADS)

Lafonteese, David James

This thesis explores the diagnostic possibilities of a fast wave-based method for measuring the ion density and temperature profiles of tokamak plasmas. In these studies fast waves are coupled to the plasma at frequencies at the second harmonic of the ion gyrofrequency, at which wave energy is absorbed by the finite-temperature ions. As the ion gyrofrequency is dependent upon the local magnetic field, which varies as l/R in a tokamak, this power absorption is radially localized. The simultaneous launching of multiple frequencies, all resonating at different plasma positions, allows local measurements of the ion density and temperature. To investigate the profile applications of wave damping measurements in a simulated tokamak, an inhouse slab-model ICRF code is developed. A variety of analysis methods are presented, and ion density and temperature profiles are reconstructed for hydrogen plasmas for the Electric Tokamak (ET) and ITER parameter spaces. These methods achieve promising results in simulated plasmas featuring bulk ion heating, off-axis RF heating, and density ramps. The experimental results of similar studies on the Electric Tokamak, a high aspect ratio (R/a = 5), low toroidal field (2.2 kG) device are then presented. In these studies, six fast wave frequencies were coupled using a single-strap, low-field-side antenna to ET plasmas. The frequencies were variable, and could be tuned to resonate at different radii for different experiments. Four magnetic pickup loops were used to measure of the toroidal component of the wave magnetic field. The expected greater eigenmode damping of center-resonant frequencies versus edge-resonant frequencies is consistently observed. Comparison of measured aspects of fast wave behavior in ET is made with the slab code predictions, which validate the code simulations under weakly-damped conditions. A density profile is measured for an ET discharge through analysis of the fast wave measurements, and is compared to an electron density profile derived from Thomson scattering data. The methodology behind a similar measurement of the ion temperature profile is also presented.

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.

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%.

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.

Simulation of neoclassical transport with the continuum gyrokinetic code COGENT

DOE PAGES

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

2013-01-25

The development of the continuum gyrokinetic code COGENT for edge plasma simulations is reported. The present version of the code models a nonlinear axisymmetric 4D (R, v∥, μ) gyrokinetic equation coupled to the long-wavelength limit of the gyro-Poisson equation. Here, R is the particle gyrocenter coordinate in the poloidal plane, and v∥ and μ are the guiding center velocity parallel to the magnetic field and the magnetic moment, respectively. The COGENT code utilizes a fourth-order finite-volume (conservative) discretization combined with arbitrary mapped multiblock grid technology (nearly field-aligned on blocks) to handle the complexity of tokamak divertor geometry with high accuracy.more » Furthermore, topics presented are the implementation of increasingly detailed model collision operators, and the results of neoclassical transport simulations including the effects of a strong radial electric field characteristic of a tokamak pedestal under H-mode conditions.« less

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

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

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

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.

3D-DIVIMP-HC modeling analysis of methane injection into DIII-D using the DiMES porous plug injector

NASA Astrophysics Data System (ADS)

Mu, Y.; McLean, A. G.; Elder, J. D.; Stangeby, P. C.; Bray, B. D.; Brooks, N. H.; Davis, J. W.; Fenstermacher, M. E.; Groth, M.; Lasnier, C. J.; Rudakov, D. L.; Watkins, J. G.; West, W. P.; Wong, C. P. C.

2009-06-01

A self-contained gas injection system for the Divertor Material Evaluation System (DiMES) on DIII-D, the porous plug injector (PPI), has been employed for in situ study of chemical erosion in the tokamak divertor environment by injection of CH 4 [A.G. McLean et al., these Proceedings]. A new interpretive code, 3D-DIVIMP-HC, has been developed and applied to the interpretation of the CH, CI, and CII emissions. Particular emphasis is placed on the interpretation of 2D filtered-camera (TV) pictures in CH, CI and CII light taken from a view essentially straight down on the PPI. The code replicates sufficient measurements to conclude that most of the basic elements of the controlling physics and chemistry have been identified and incorporated in the code-model.

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.

Implementing a finite-state off-normal and fault response system for disruption avoidance in tokamaks

NASA Astrophysics Data System (ADS)

Eidietis, N. W.; Choi, W.; Hahn, S. H.; Humphreys, D. A.; Sammuli, B. S.; Walker, M. L.

2018-05-01

A finite-state off-normal and fault response (ONFR) system is presented that provides the supervisory logic for comprehensive disruption avoidance and machine protection in tokamaks. Robust event handling is critical for ITER and future large tokamaks, where plasma parameters will necessarily approach stability limits and many systems will operate near their engineering limits. Events can be classified as off-normal plasmas events, e.g. neoclassical tearing modes or vertical displacements events, or faults, e.g. coil power supply failures. The ONFR system presented provides four critical features of a robust event handling system: sequential responses to cascading events, event recovery, simultaneous handling of multiple events and actuator prioritization. The finite-state logic is implemented in Matlab®/Stateflow® to allow rapid development and testing in an easily understood graphical format before automated export to the real-time plasma control system code. Experimental demonstrations of the ONFR algorithm on the DIII-D and KSTAR tokamaks are presented. In the most complex demonstration, the ONFR algorithm asynchronously applies ‘catch and subdue’ electron cyclotron current drive (ECCD) injection scheme to suppress a virulent 2/1 neoclassical tearing mode, subsequently shuts down ECCD for machine protection when the plasma becomes over-dense, and enables rotating 3D field entrainment of the ensuing locked mode to allow a safe rampdown, all in the same discharge without user intervention. When multiple ONFR states are active simultaneously and requesting the same actuator (e.g. neutral beam injection or gyrotrons), actuator prioritization is accomplished by sorting the pre-assigned priority values of each active ONFR state and giving complete control of the actuator to the state with highest priority. This early experience makes evident that additional research is required to develop an improved actuator sharing protocol, as well as a methodology to minimize the number and topological complexity of states as the finite-state ONFR system is scaled to a large, highly constrained device like ITER.

Implementing a finite-state off-normal and fault response system for disruption avoidance in tokamaks

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

Eidietis, N. W.; Choi, W.; Hahn, S. H.

A finite-state off-normal and fault response (ONFR) system is presented that provides the supervisory logic for comprehensive disruption avoidance and machine protection in tokamaks. Robust event handling is critical for ITER and future large tokamaks, where plasma parameters will necessarily approach stability limits and many systems will operate near their engineering limits. Events can be classified as off-normal plasmas events, e.g. neoclassical tearing modes or vertical displacements events, or faults, e.g. coil power supply failures. The ONFR system presented provides four critical features of a robust event handling system: sequential responses to cascading events, event recovery, simultaneous handling of multiplemore » events and actuator prioritization. The finite-state logic is implemented in Matlab*/Stateflow* to allow rapid development and testing in an easily understood graphical format before automated export to the real-time plasma control system code. Experimental demonstrations of the ONFR algorithm on the DIII-D and KSTAR tokamaks are presented. In the most complex demonstration, the ONFR algorithm asynchronously applies “catch and subdue” electron cyclotron current drive (ECCD) injection scheme to suppress a virulent 2/1 neoclassical tearing mode, subsequently shuts down ECCD for machine protection when the plasma becomes over-dense, and enables rotating 3D field entrainment of the ensuing locked mode to allow a safe rampdown, all in the same discharge without user intervention. When multiple ONFR states are active simultaneously and requesting the same actuator (e.g. neutral beam injection or gyrotrons), actuator prioritization is accomplished by sorting the pre-assigned priority values of each active ONFR state and giving complete control of the actuator to the state with highest priority. This early experience makes evident that additional research is required to develop an improved actuator sharing protocol, as well as a methodology to minimize the number and topological complexity of states as the finite-state ONFR system is scaled to a large, highly constrained device like ITER.« less

Implementing a finite-state off-normal and fault response system for disruption avoidance in tokamaks

DOE PAGES

Eidietis, N. W.; Choi, W.; Hahn, S. H.; ...

2018-03-29

A finite-state off-normal and fault response (ONFR) system is presented that provides the supervisory logic for comprehensive disruption avoidance and machine protection in tokamaks. Robust event handling is critical for ITER and future large tokamaks, where plasma parameters will necessarily approach stability limits and many systems will operate near their engineering limits. Events can be classified as off-normal plasmas events, e.g. neoclassical tearing modes or vertical displacements events, or faults, e.g. coil power supply failures. The ONFR system presented provides four critical features of a robust event handling system: sequential responses to cascading events, event recovery, simultaneous handling of multiplemore » events and actuator prioritization. The finite-state logic is implemented in Matlab*/Stateflow* to allow rapid development and testing in an easily understood graphical format before automated export to the real-time plasma control system code. Experimental demonstrations of the ONFR algorithm on the DIII-D and KSTAR tokamaks are presented. In the most complex demonstration, the ONFR algorithm asynchronously applies “catch and subdue” electron cyclotron current drive (ECCD) injection scheme to suppress a virulent 2/1 neoclassical tearing mode, subsequently shuts down ECCD for machine protection when the plasma becomes over-dense, and enables rotating 3D field entrainment of the ensuing locked mode to allow a safe rampdown, all in the same discharge without user intervention. When multiple ONFR states are active simultaneously and requesting the same actuator (e.g. neutral beam injection or gyrotrons), actuator prioritization is accomplished by sorting the pre-assigned priority values of each active ONFR state and giving complete control of the actuator to the state with highest priority. This early experience makes evident that additional research is required to develop an improved actuator sharing protocol, as well as a methodology to minimize the number and topological complexity of states as the finite-state ONFR system is scaled to a large, highly constrained device like ITER.« less

Fuel cycle for a fusion neutron source

NASA Astrophysics Data System (ADS)

Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

2015-12-01

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

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.

Dynamical coupling between magnetic equilibrium and transport in tokamak scenario modelling, with application to current ramps

NASA Astrophysics Data System (ADS)

Fable, E.; Angioni, C.; Ivanov, A. A.; Lackner, K.; Maj, O.; Medvedev, S. Yu; Pautasso, G.; Pereverzev, G. V.; Treutterer, W.; the ASDEX Upgrade Team

2013-07-01

The modelling of tokamak scenarios requires the simultaneous solution of both the time evolution of the plasma kinetic profiles and of the magnetic equilibrium. Their dynamical coupling involves additional complications, which are not present when the two physical problems are solved separately. Difficulties arise in maintaining consistency in the time evolution among quantities which appear in both the transport and the Grad-Shafranov equations, specifically the poloidal and toroidal magnetic fluxes as a function of each other and of the geometry. The required consistency can be obtained by means of iteration cycles, which are performed outside the equilibrium code and which can have different convergence properties depending on the chosen numerical scheme. When these external iterations are performed, the stability of the coupled system becomes a concern. In contrast, if these iterations are not performed, the coupled system is numerically stable, but can become physically inconsistent. By employing a novel scheme (Fable E et al 2012 Nucl. Fusion submitted), which ensures stability and physical consistency among the same quantities that appear in both the transport and magnetic equilibrium equations, a newly developed version of the ASTRA transport code (Pereverzev G V et al 1991 IPP Report 5/42), which is coupled to the SPIDER equilibrium code (Ivanov A A et al 2005 32nd EPS Conf. on Plasma Physics (Tarragona, 27 June-1 July) vol 29C (ECA) P-5.063), in both prescribed- and free-boundary modes is presented here for the first time. The ASTRA-SPIDER coupled system is then applied to the specific study of the modelling of controlled current ramp-up in ASDEX Upgrade discharges.

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.

Recent progress of the Laser-driven Ion-beam Trace Probe

NASA Astrophysics Data System (ADS)

Yang, Xiaoyi; Xiao, Chijie; Chen, Yihang; 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 method to diagnose the poloidal magnetic field and radial electric field in tokamaks. Recently significant progresses have been made as follows. 1) The experimental system has been set up on the PKU Plasma Test (PPT) linear device and begun to validate the principle of LITP, including the ion source, the ion detector and the poloidal magnetic field cable. Preliminary experimental results matched the theoretical prediction well. 2) The reconstruction principle has been improved including the nonlinear effect. 3) Tomography methods have been applied in the reconstruction codes. Now the laser-driven ion-beam accelerator has been setup on the PPT device, and further test of LITP will start soon. After that a prototype of LITP system will be designed and setup on the HL-2A tokamak device. This work was supported by the CHINA MOST under 2012YQ030142, ITER-CHINA program 2015GB120001 and National Natural Science Foundation of China under 11575014 and 11375053.

Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

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

Du, T. F.; Chen, Z. J.; Peng, X. Y.

A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometermore » at EAST are studied for future data interpretation.« less

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

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

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.

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.

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.

Experimental studies of toroidal correlations of plasma density fluctuations along the magnetic field lines in the T-10 tokamak and first results of numerical modeling

NASA Astrophysics Data System (ADS)

Buldakov, M. A.; Vershkov, V. A.; Isaev, M. Yu; Shelukhin, D. A.

2017-10-01

The antenna system of reflectometry diagnostics at the T-10 tokamak allows to study long-range toroidal correlations of plasma density fluctuations along the magnetic field lines. The antenna systems are installed in two poloidal cross-sections of the vacuum chamber separated by a 90° angle in the toroidal direction. The experiments, which were conducted at the low field side, showed that the high level of toroidal correlations is observed only for quasi-coherent fluctuations. However, broadband and stochastic low frequency fluctuations are not correlated. Numerical modeling of the plasma turbulence structure in the T-10 tokamak was conducted to interpret the experimental results and take into account non-locality of reflectometry measurements. In the model used, it was assumed that the magnitudes of density fluctuations are constant along the magnetic field lines. The 2D full-wave Tamic-RTH code was used to model the reflectometry signals. High level of correlations for quasi-coherent fluctuations was obtained during the modeling, which agrees with the experimental observations. However, the performed modeling also predicts high level of correlations for broadband fluctuations, which contradicts the experimental data. The modeling showed that the effective reflection radius, from which the information on quasi-coherent plasma turbulence is obtained, is shifted outwards from the reflection radius by approximately 7 mm.

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.

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.

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.

Simulations of Turbulence in Tokamak Edge and Effects of Self-Consistent Zonal Flows

NASA Astrophysics Data System (ADS)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress is reported on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge. This extends previous work to include self-consistent zonal flows and their effects. The previous work addressed simulation of L-mode tokamak edge turbulence using the turbulence code BOUT that solves Braginskii-based plasma fluid equations in tokamak edge domain. The calculations use realistic single-null geometry and plasma parameters of the DIII-D tokamak and produce 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 U.S. Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

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.

Simulation of magnetic island dynamics under resonant magnetic perturbation with the TEAR code and validation of the results on T-10 tokamak data

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

Ivanov, N. V.; Kakurin, A. M.

2014-10-15

Simulation of the magnetic island evolution under Resonant Magnetic Perturbation (RMP) in rotating T-10 tokamak plasma is presented with intent of TEAR code experimental validation. In the T-10 experiment chosen for simulation, the RMP consists of a stationary error field, a magnetic field of the eddy current in the resistive vacuum vessel and magnetic field of the externally applied controlled halo current in the plasma scrape-off layer (SOL). The halo-current loop consists of a rail limiter, plasma SOL, vacuum vessel, and external part of the circuit. Effects of plasma resistivity, viscosity, and RMP are taken into account in the TEARmore » code based on the two-fluid MHD approximation. Radial distribution of the magnetic flux perturbation is calculated with account of the externally applied RMP. A good agreement is obtained between the simulation results and experimental data for the cases of preprogrammed and feedback-controlled halo current in the plasma SOL.« less

Protection of tokamak plasma facing components by a capillary porous system with lithium

NASA Astrophysics Data System (ADS)

Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.

2015-08-01

Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

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.

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

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

Harvey, R. W.

This DOE grant supported fusion energy research, a potential long-term solution to the world's energy needs. Magnetic fusion, exemplified by confinement of very hot ionized gases, i.e., plasmas, in donut-shaped tokamak vessels is a leading approach for this energy source. Thus far, a mixture of hydrogen isotopes has produced 10's of megawatts of fusion power for seconds in a tokamak reactor at Princeton Plasma Physics Laboratory in New Jersey. The research grant under consideration, ER54684, uses computer models to aid in understanding and projecting efficacy of heating and current drive sources in the National Spherical Torus Experiment, a tokamak variant,more » at PPPL. The NSTX experiment explores the physics of very tight aspect ratio, almost spherical tokamaks, aiming at producing steady-state fusion plasmas. The current drive is an integral part of the steady-state concept, maintaining the magnetic geometry in the steady-state tokamak. CompX further developed and applied models for radiofrequency (rf) heating and current drive for applications to NSTX. These models build on a 30 year development of rf ray tracing (the all-frequencies GENRAY code) and higher dimensional Fokker-Planck rf-collisional modeling (the 3D collisional-quasilinear CQL3D code) at CompX. Two mainline current-drive rf modes are proposed for injection into NSTX: (1) electron Bernstein wave (EBW), and (2) high harmonic fast wave (HHFW) modes. Both these current drive systems provide a means for the rf to access the especially high density plasma--termed high beta plasma--compared to the strength of the required magnetic fields. The CompX studies entailed detailed modeling of the EBW to calculate the efficiency of the current drive system, and to determine its range of flexibility for driving current at spatial locations in the plasma cross-section. The ray tracing showed penetration into NSTX bulk plasma, relatively efficient current drive, but a limited ability to produce current over the whole radial plasma cross-section. The actual EBW experiment will cost several million dollars, and remains in the proposal stage. The HHFW current drive system has been experimentally implemented on NSTX, and successfully drives substantial current. The understanding of the experiment is to be accomplished in terms of general concepts of rf current drive, and also detailed modeling of the experiment which can discern the various competing processes which necessarily occur simultaneously in the experiment. An early discovery of the CompX codes, GENRAY and CQL3D, was that there could be significant interference between the neutral beam injection fast ions in the machine (injected for plasma heating) and the HHFW energy. Under many NSTX experimental conditions, power which could go to the fast ions would then be unavailable for current drive by the desired HHFW interaction with electrons. This result has been born out by experiments; the modeling helps in understanding difficulties with HHFW current drive, and has enabled adjustment of the experiment to avoid interaction with neutral beam injected fast ions thereby achieving stronger HHFW current drive. The detailed physics modeling of the various competing processes is almost always required in fusion energy plasma physics, to ensure a reasonably accurate and certain interpretation of the experiment, enabling the confident design of future, more advanced experiments and ultimately a commercial fusion reactor. More recent work entails detailed investigation of the interaction of the HHFW radiation for fast ions, accounting for the particularly large radius orbits in NSTX, and correlations between multiple HHFW-ion interactions. The spherical aspect of the NSTX experiment emphasized particular physics such as the large orbits which are present to some degree in all tokamaks, but gives clearer clues on the resulting physics phenomena since competing physics effects are reduced.« less

3D-DIVIMP(HC) code modeling of DIII-D DiMES porous plug injector experiments

NASA Astrophysics Data System (ADS)

Mu, Y.; Elder, J. D.; Stangeby, P. C.; McLean, A. G.

2011-08-01

A Porous Plug Injector (PPI) system for the Divertor Material Evaluation System (DiMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. The 3D-DIVIMP(HC) code has been applied to the interpretation of the CI, CII and other spectroscopic measurements made at the PPI location, for (a) the synthetic source due to injection of CH4 through the PPI, and (b) the natural emission from the PPI head itself, which was inserted above surrounding graphite tiles by ˜0.3 mm.The code successfully replicated the MDS (spectrometer)-measured absolute emissions of CH, CI, CII 427 nm, 514 nm, and 658 nm [1] and the DiMES TV-measured spatial shapes of the CH, CI, and CII 514 nm [1] emission "clouds" to within the combined uncertainties. It is thus concluded that the most important physics and chemistry of chemical sputtering have most likely been included in the model.

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.

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.

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.

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.

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.

Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks

NASA Astrophysics Data System (ADS)

Peysson, Y.; Bonoli, P. T.; Chen, J.; Garofalo, A.; Hillairet, J.; Li, M.; Qian, J.; Shiraiwa, S.; Decker, J.; Ding, B. J.; Ekedahl, A.; Goniche, M.; Zhai, X.

2017-10-01

The Lower Hybrid (LH) wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum). Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model), for robust interpretative and predictive simulations.

ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

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

Berry, Jan; Ferrada, Juan J; Curd, Warren

During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predictedmore » to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support from Northrop Grumman, and OneCIS). ITER International Organization (ITER-IO) is responsible for design oversight and equipment installation in Cadarache, France. TCWS equipment will be fabricated using ASME design codes with quality assurance and oversight by an Agreed Notified Body (approved by the French regulator) that will ensure regulatory compliance. This paper describes the TCWS design and how U.S. ITER and fabricators will use ASME codes to comply with EU Directives and French Orders and Decrees.« less

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

Radial and poloidal correlation reflectometry on Experimental Advanced Superconducting Tokamak

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

Qu, Hao; Zhang, Tao; Han, Xiang

2015-08-15

An X-mode polarized V band (50 GHz–75 GHz) radial and poloidal correlation reflectometry is designed and installed on Experimental Advanced Superconducting Tokamak (EAST). Two frequency synthesizers (12 GHz–19 GHz) are used as sources. Signals from the sources are up-converted to V band using active quadruplers and then coupled together for launching through one single pyramidal antenna. Two poloidally separated antennae are installed to receive the reflected waves from plasma. This reflectometry system can be used for radial and poloidal correlation measurement of the electron density fluctuation. In ohmically heated plasma, the radial correlation length is about 1.5 cm measured bymore » the system. The poloidal correlation analysis provides a means to estimate the fluctuation velocity perpendicular to the main magnetic field. In the present paper, the distance between two poloidal probing points is calculated with ray-tracing code and the propagation time is deduced from cross-phase spectrum. Fluctuation velocity perpendicular to the main magnetic field in the core of ohmically heated plasma is about from −1 km/s to −3 km/s.« 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.

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

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.

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.

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.

Improvements to the National Transport Code Collaboration Data Server

NASA Astrophysics Data System (ADS)

Alexander, David A.

2001-10-01

The data server of the National Transport Code Colaboration Project provides a universal network interface to interpolated or raw transport data accessible by a universal set of names. Data can be acquired from a local copy of the Iternational Multi-Tokamak (ITER) profile database as well as from TRANSP trees of MDS Plus data systems on the net. Data is provided to the user's network client via a CORBA interface, thus providing stateful data server instances, which have the advantage of remembering the desired interpolation, data set, etc. This paper will review the status and discuss the recent improvements made to the data server, such as the modularization of the data server and the addition of hdf5 and MDS Plus data file writing capability.

Modelling the power deposition into a spherical tokamak fusion power plant

NASA Astrophysics Data System (ADS)

Windsor, C. G.; Morgan, J. G.; Buxton, P. F.; Costley, A. E.; Smith, G. D. W.; Sykes, A.

2017-03-01

Numerical studies have been made to improve the performance of the central column of a superconducting spherical tokamak fusion pilot plant. The assumed neutron shield includes concentric layers of tungsten carbide and water. The relative thickness of the water layers was varied and a minimum power deposition was found at about 17% of water. It was found advantageous to have an approximately 1.7 times thicker water layer next to the core and a similarly thinner layer next to the plasma. The use of tungsten boride instead of tungsten carbide was shown to make an improvement especially if placed close to the central superconducting core, the inner layer alone reducing the power deposition by 29%. Engineering features such as a central steel tie-bar, an insulating thermal vacuum gap, a wall gap next to the plasma and knowledge of the vertical energy distribution are essential to a successful design and their effects on the power deposition are shown in an appendix. The results have been fitted to model distributions and incorporated into the Tokamak Energy System Code, which can then give predictions of the power deposition as a function of other parameters such as the plasma major radius and the maximum magnetic field permitted on the superconductors.

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.

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.

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.

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.

Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor

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

Kim, Seokho H; Berry, Jan

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclearmore » pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.« 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.

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.

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.

Integrated Design of Undepressed Collector for Low Power Gyrotron

NASA Astrophysics Data System (ADS)

Kumar, Anil; Goswami, Uttam K.; Poonia, Sunita; Singh, Udaybir; Kumar, Nitin; Alaria, M. K.; Bera, A.; Khatun, Hasina; Sinha, A. K.

2011-06-01

A 42 GHz, 200 kW continuous wave (CW) gyrotron, operating at TE03 mode is under development for the electron cyclotron resonance plasma heating of the Indian TOKAMAK system. The gyrotron is made up of an undepressed collector. The undepressed collector is simple to design and cost effective. In this paper, a detailed design study of the undepressed collector for the 42 GHz gyrotron is presented. The EGUN code is used to analyze the spent electron beam trajectory for the maximum spread to reduce the power loading on the collector surface. To achieve wall loading ≤1 kW/cm2, a collector with a length of 800 mm and a radius of 42.5 mm is designed. The design also includes the three magnet systems around the collector for maximum and uniform beam spread. The thermal and the structural analyses are done using the ANSYS code to optimize the collector structure and dimensions with tolerance.

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.

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.

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.

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

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.

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.

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.

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

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

Study of runaway electrons in TUMAN-3M tokamak plasmas

NASA Astrophysics Data System (ADS)

Shevelev, A.; Khilkevitch, E.; Tukachinsky, A.; Pandya, S.; Askinazi, L.; Belokurov, A.; Chugunov, I.; Doinikov, D.; Gin, D.; Iliasova, M.; Kiptily, V.; Kornev, V.; Lebedev, S.; Naidenov, V.; Plyusnin, V.; Polunovsky, I.; Zhubr, N.

2018-07-01

Studies of runaway electrons in present day tokamaks are essential to improve theoretical models and to support possible avoidance or suppression mechanisms in future large-scale plasma devices. Some of the phenomena associated with the runaway electrons take place at faster time scales, and thus it is essential to probe the runaway electrons to investigate underlying physics. The present article reports a few experimental observations of runaway electron associated events, at fast time scales, using a state-of-the-art multi-detector system developed at the Ioffe Institute and recently deployed on the TUMAN-3M tokamak. The system is based on the high-performance scintillation gamma-ray spectrometers for measurements of bremsstrahlung generated during the interaction of accelerated electrons with plasma and materials of the tokamak chamber. It includes a total three detectors configured in the spectroscopic mode having different lines of sight. Along with this hardware, dedicated algorithms were developed and validated that enables the separation of piled-up pulses, maximize the dynamic range of the detector and provides a counting rate as high as 107 counts per second. The inversion code, DeGaSum, has been used for the reconstruction of a runaway electron energy distribution function from the measured gamma-ray spectra. Using this tool, experimental analysis of the runaway electron beam generation and evolution of their energy distribution in the TUMAN-3M representative plasma discharges is performed. The effect on gamma-ray count rate during the magnetohydrodynamic activities and possible changes in the runaway electron energy distribution function during sawtooth oscillations is discussed in detail. Possible maximum limit of the runaway electron energy in TUMAN-3M is investigated and compared with the numerical analysis. In addition, the probability of the runaway electron generation throughout the plasma discharge is estimated analytically and compared with the experimental observation that suggests a balance between production and loss of the runaway electrons.

NSTX-U Control System Upgrades

DOE PAGES

Erickson, K. G.; Gates, D. A.; Gerhardt, S. P.; ...

2014-06-01

The National Spherical Tokamak Experiment (NSTX) is undergoing a wealth of upgrades (NSTX-U). These upgrades, especially including an elongated pulse length, require broad changes to the control system that has served NSTX well. A new fiber serial Front Panel Data Port input and output (I/O) stream will supersede the aging copper parallel version. Driver support for the new I/O and cyber security concerns require updating the operating system from Redhat Enterprise Linux (RHEL) v4 to RedHawk (based on RHEL) v6. While the basic control system continues to use the General Atomics Plasma Control System (GA PCS), the effort to forwardmore » port the entire software package to run under 64-bit Linux instead of 32-bit Linux included PCS modifications subsequently shared with GA and other PCS users. Software updates focused on three key areas: (1) code modernization through coding standards (C99/C11), (2) code portability and maintainability through use of the GA PCS code generator, and (3) support of 64-bit platforms. Central to the control system upgrade is the use of a complete real time (RT) Linux platform provided by Concurrent Computer Corporation, consisting of a computer (iHawk), an operating system and drivers (RedHawk), and RT tools (NightStar). Strong vendor support coupled with an extensive RT toolset influenced this decision. The new real-time Linux platform, I/O, and software engineering will foster enhanced capability and performance for NSTX-U plasma control.« less

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.

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.

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

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.

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

Ekedahl, Annika, E-mail: annika.ekedahl@cea.fr; Bourdelle, Clarisse; Artaud, Jean-François

The longstanding expertise of the Tore Supra team in long pulse heating and current drive with radiofrequency (RF) systems will now be exploited in the WEST device (tungsten-W Environment in Steady-state Tokamak) [1]. WEST will allow an integrated long pulse tokamak programme for testing W-divertor components at ITER-relevant heat flux (10-20 MW/m{sup 2}), while treating crucial aspects for ITER-operation, such as avoidance of W-accumulation in long discharges, monitoring and control of heat fluxes on the metallic plasma facing components (PFCs) and coupling of RF waves in H-mode plasmas. Scenario modelling using the METIS-code shows that ITER-relevant heat fluxes are compatiblemore » with the sustainment of long pulse H-mode discharges, at high power (up to 15 MW / 30 s at I{sub P} = 0.8 MA) or high fluence (up to 10 MW / 1000 s at I{sub P} = 0.6 MA) [2], all based on RF heating and current drive using Ion Cyclotron Resonance Heating (ICRH) and Lower Hybrid Current Drive (LHCD). This paper gives a description of the ICRH and LHCD systems in WEST, together with the modelling of the power deposition of the RF waves in the WEST-scenarios.« less

ITER Port Interspace Pressure Calculations

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

Carbajo, Juan J; Van Hove, Walter A

The ITER Vacuum Vessel (VV) is equipped with 54 access ports. Each of these ports has an opening in the bioshield that communicates with a dedicated port cell. During Tokamak operation, the bioshield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a Port Interspace (between VV closure lid and Port Plug) on the inner side and the Port Cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER (Ref. 1) Port Interspace after a double-ended guillotine break (DEGB)more » of a pipe of the Tokamak Cooling Water System (TCWS) with high temperature water. It is assumed that this DEGB occurs during the worst possible conditions, which are during water baking operation, with water at a temperature of 523 K (250 C) and at a pressure of 4.4 MPa. These conditions are more severe than during normal Tokamak operation, with the water at 398 K (125 C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 (Ref. 2) to calculate the blowdown releases from the pipe break, and MELCOR, Version 1.8.6 (Ref. 3) to calculate the pressures and temperatures in the Port Interspace. A sensitivity study has been performed to optimize some flow areas.« less

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

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

Jackson, G. L.; Politzer, P. A.; Humphreys, D. A.

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

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.

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

Protecting Against Damage from Refraction of High Power Microwaves in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Lohr, John; Brambila, Rigo; Cengher, Mirela; Chen, Xi; Gorelov, Yuri; Grosnickle, William; Moeller, Charles; Ponce, Dan; Prater, Ron; Torrezan, Antonio; Austin, Max; Doyle, Edward; Hu, Xing; Dormier, Calvin

2017-07-01

Several new protective systems are being installed on the DIII D tokamak to increase the safety margins for plasma operations with injected ECH power at densities approaching cutoff. Inadvertent overdense operation has previously resulted in reflection of an rf beam back into a launcher causing extensive arcing and melt damage on one waveguide line. Damage to microwave diagnostics, which are located on the same side of the tokamak as the ECH launchers, also has occurred. Developing a reliable microwave based interlock to protect the many vulnerable systems in DIII-D has proved to be difficult. Therefore, multiple protective steps have been taken to reduce the risk of damage in the future. Among these is a density interlock generated by the plasma control system, with setpoint determined by the ECH operators based on rf beam trajectories and plasma parameters. Also installed are enhanced video monitoring of the launchers, and an ambient light monitor on each of the waveguide systems, along with a Langmuir probe at the mouth of each launcher. Versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, have been installed as the last miter bends in each waveguide line. As these systems are characterized, they are being incorporated in the interlock chains, which enable the ECH injection permits. The diagnostics most susceptible to damage from the ECH waves have also been fitted with a variety of protective devices including stripline filters, thin resonant notch filters tuned to the 110 GHz injected microwave frequency, blazed grating filters and shutters. Calculations of rf beam trajectories in the plasmas are performed using the TORAY ray tracing code with input from kinetic profile diagnostics. Using these calculations, strike points for refracted beams on the vacuum vessel are calculated, which allows evaluation of the risk of damage to sensitive diagnostics and hardware.

Protecting against damage from refraction of high power microwaves in the DIII-D tokamak

DOE PAGES

Lohr, John; Brambila, Rigo; Cengher, Mirela; ...

2017-07-24

Here, several new protective systems are being installed on the DIII D tokamak to increase the safety margins for plasma operations with injected ECH power at densities approaching cutoff. Inadvertent overdense operation has previously resulted in reflection of an rf beam back into a launcher causing extensive arcing and melt damage on one waveguide line. Damage to microwave diagnostics, which are located on the same side of the tokamak as the ECH launchers, also has occurred. Developing a reliable microwave based interlock to protect the many vulnerable systems in DIII-D has proved to be difficult. Therefore, multiple protective steps havemore » been taken to reduce the risk of damage in the future. Among these is a density interlock generated by the plasma control system, with setpoint determined by the ECH operators based on rf beam trajectories and plasma parameters. Also installed are enhanced video monitoring of the launchers, and an ambient light monitor on each of the waveguide systems, along with a Langmuir probe at the mouth of each launcher. Versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, have been installed as the last miter bends in each waveguide line. As these systems are characterized, they are being incorporated in the interlock chains, which enable the ECH injection permits. The diagnostics most susceptible to damage from the ECH waves have also been fitted with a variety of protective devices including stripline filters, thin resonant notch filters tuned to the 110 GHz injected microwave frequency, blazed grating filters and shutters. Calculations of rf beam trajectories in the plasmas are performed using the TORAY ray tracing code with input from kinetic profile diagnostics. Using these calculations, strike points for refracted beams on the vacuum vessel are calculated, which allows evaluation of the risk of damage to sensitive diagnostics and hardware.« less

Protecting against damage from refraction of high power microwaves in the DIII-D tokamak

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

Lohr, John; Brambila, Rigo; Cengher, Mirela

Here, several new protective systems are being installed on the DIII D tokamak to increase the safety margins for plasma operations with injected ECH power at densities approaching cutoff. Inadvertent overdense operation has previously resulted in reflection of an rf beam back into a launcher causing extensive arcing and melt damage on one waveguide line. Damage to microwave diagnostics, which are located on the same side of the tokamak as the ECH launchers, also has occurred. Developing a reliable microwave based interlock to protect the many vulnerable systems in DIII-D has proved to be difficult. Therefore, multiple protective steps havemore » been taken to reduce the risk of damage in the future. Among these is a density interlock generated by the plasma control system, with setpoint determined by the ECH operators based on rf beam trajectories and plasma parameters. Also installed are enhanced video monitoring of the launchers, and an ambient light monitor on each of the waveguide systems, along with a Langmuir probe at the mouth of each launcher. Versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, have been installed as the last miter bends in each waveguide line. As these systems are characterized, they are being incorporated in the interlock chains, which enable the ECH injection permits. The diagnostics most susceptible to damage from the ECH waves have also been fitted with a variety of protective devices including stripline filters, thin resonant notch filters tuned to the 110 GHz injected microwave frequency, blazed grating filters and shutters. Calculations of rf beam trajectories in the plasmas are performed using the TORAY ray tracing code with input from kinetic profile diagnostics. Using these calculations, strike points for refracted beams on the vacuum vessel are calculated, which allows evaluation of the risk of damage to sensitive diagnostics and hardware.« less

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)

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

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.

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.

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.

Electron Temperature Fluctuation Measurements and Transport Model Validation at Alcator C-Mod

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

White, Anne

The tokamak is a type of toroidal device used to confine a fusion plasma using large magnetic fields. Tokamaks and stellarators the leading devices for confining plasmas for fusion, and the capability to predict performance in these magnetically confined plasmas is essential for developing a sustainable fusion energy source. The magnetic configuration of tokamaks and stellarators does not exist in Nature, yet, the fundamental processes governing transport in fusion plasmas are universal – turbulence and instabilities, driven by inhomogeneity and asymmetry in the plasma, conspire to transport heat and particles across magnetic field lines and can play critical roles inmore » impurity confinement and generation of intrinsic rotation. Turbulence exists in all plasmas, and in neutral fluids as well. The study of turbulence is essential to developing a fundamental understanding of the nature of the fourth state of matter, plasmas. Experimental studies of turbulence in tokamaks date back to early scattering observations from the late 1970s. Since that time, great advances in turbulence diagnostics have been made, all of which have significantly enhanced our knowledge and understanding of turbulence in tokamaks. Through comparisons with advanced gyrokinetic theory and turbulent-transport models a great deal of evidence exists to implicate turbulent-driven transport as an important mechanism determining transport in all channels: heat, particle and momentum However, prediction and control of turbulent-driven transport remains elusive. Key to development of predictive transport models for magnetically confined fusion plasmas is validation of the nonlinear gyrokinetic transport model, which describes transport due to turbulence. Validation of gyrokinetic codes must include detailed and quantitative comparisons with measured turbulence characteristics, in addition to comparisons with inferred transport levels and equilibrium profiles. For this reason, advanced plasma diagnostics for studying core turbulence are needed in order to assess the accuracy of gyrokinetic models for turbulent-driven particle, heat and momentum transport. New core turbulence diagnostics at the world-class tokamaks Alcator C-Mod at MIT and ASDEX Upgrade at the Max Planck Institute for Plasma Physics have been designed, developed, and operated over the course of this project. These new instruments are capable of measuring electron temperature fluctuations and the phase angle between density and temperature fluctuations locally and quantitatively. These new data sets from Alcator C-Mod and ASDEX Upgrade are being used to fill key gaps in our understanding of turbulent transport in tokamaks. In particular, this project has results in new results on the topics of the Transport Shortfall, the role of ETG turbulence in tokamak plasmas, profile stiffness, the LOC/SOC transition, and intrinsic rotation reversals. These data are used in a rigorous process of “Transport model validation”, and this group is a world-leader on using turbulence models to design new hardware and new experiments at tokamaks. A correlation electron cyclotron emission (CECE) diagnostic is an instrument used to measure micro-scale fluctuations (mm-scale, compared to the machine size of meters) of electron temperature in magnetically confined fusion plasmas, such as those in tokamaks and stellarators. These micro-scale fluctuations are associated with drift-wave type turbulence, which leads to enhanced cooling and mixing of particles in fusion plasmas and limits achieving the required temperatures and densities for self-sustained fusion reactions. A CECE system can also be coupled with a reflectometer system that measured micro-scale density fluctuations, and from these simultaneous measurements, one can extract the phase between the density (n) and temperature (T) fluctuations, creating an nT phase diagnostic. Measurements of the fluctuations and the phase angle between them are extremely useful for testing and validating predictive models for the transport of heat and particles in fusion plasmas due to turbulence. Once validated, the models are used to predict performance in ITER and other burning plasmas, such as the MIT ARC design. Most recently, data from the newly developed, so-called “CECE diagnostic” [Cima 1995, White 2008] and “nT phase angle measurements” [Haese 1999, White 2010] ]will be combined with data from density fluctuation diagnostics at ASDEX Upgrade to support a long-term program of physics research in turbulence and transport that will allow for more stringent testing and validation of gyrokinetic turbulent-transport codes. This work directly impacts the development of predictive transport models in the U.S. FES program, such as TGLF, developed by General Atomics, which are used to predict performance in ITER and other burning plasma devices as part of advancing the development of fusion energy sciences.« less

Sub-Alfvénic reduced magnetohydrodynamic equations for tokamaks

NASA Astrophysics Data System (ADS)

Sengupta, W.; Hassam, A. B.; Antonsen, T. M.

2017-06-01

A reduced set of magnetohydrodynamic (MHD) equations is derived, applicable to large aspect ratio tokamaks and relevant for dynamics that is sub-Alfvénic with respect to ideal ballooning modes. This ordering optimally allows sound waves, Mercier modes, drift modes, geodesic-acoustic modes (GAM), zonal flows and shear Alfvén waves. Wavelengths long compared to the gyroradius but comparable to the minor radius of a typical tokamak are considered. With the inclusion of resistivity, tearing modes, resistive ballooning modes, Pfirsch-Schluter cells and the Stringer spin-up are also included. A major advantage is that the resulting system is two-dimensional in space, and the system incorporates self-consistent and dynamic Shafranov shifts. A limitation is that the system is valid only in radial domains where the tokamak safety factor, , is close to rational. In the tokamak core, the system is well suited to study the sawtooth discharge in the presence of Mercier modes. The systematic ordering scheme and methodology developed are versatile enough to reduce the more general collisional two-fluid equations or possibly the Vlasov-Maxwell system in the MHD ordering.

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.

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.

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

1995-03-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

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

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

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

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.

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

Development and applications of 3D-DIVIMP(HC) Monte Carlo impurity modeling code

NASA Astrophysics Data System (ADS)

Mu, Yarong

A self-contained gas injection system for the Divertor Material Evaluation System (DiMES) on DIII-D, the Porous Plug Injector (PPI), has been employed by A. McLean for in-situ study of chemical erosion in the tokamak divertor environment by injection of CH4. The principal contribution of the present thesis is a new interpretive code, 3D-DIVIMP(HC), which has been developed and successfully applied to the interpretation of the CH, C I, and C II emissions measured during the PPI experiments. The two principal types of experimental data which are compared here with 3D-DIVIMP(HC) code modeling are (a) absolute emissivities measured with a high resolution spectrometer, and (b) 2D filtered camera (TV) pictures taken from a view essentially straight down on the PPI. Incorporating the Janev-Reiter database for the breakup reactions of methane molecules in a plasma, 3D-DIVIMP(HC) is able to replicate these measurements to within the combined experimental and database uncertainties. It is therefore concluded that the basic elements of the physics and chemistry controlling the breakup of methane entering an attached divertor plasma have been identified and are incorporated in 3D-DIVIMP(HC).

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.

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

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.

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

NASA Astrophysics Data System (ADS)

Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

2015-02-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

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.

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

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).

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.

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.

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.

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.

Development of a real time magnetic island identification system for HL-2A tokamak.

PubMed

Chen, Chao; Sun, Shan; Ji, Xiaoquan; Yin, Zejie

2017-08-01

A novel real time magnetic island identification system for HL-2A is introduced. The identification method is based on the measurement of Mirnov probes and the equilibrium flux constructed by the equilibrium fit (EFIT) code. The system consists of an analog front board and a digital processing board connected by a shield cable. Four octal-channel analog-to-digital convertors are utilized for 100 KHz simultaneous sampling of all the probes, and the applications of PCI extensions for Instrumentation platform and reflective memory allow the system to receive EFIT results simultaneously. A high performance field programmable gate array (FPGA) is used to realize the real time identification algorithm. Based on the parallel and pipeline processing of the FPGA, the magnetic island structure can be identified with a cycle time of 3 ms during experiments.

Development of a real time magnetic island identification system for HL-2A tokamak

NASA Astrophysics Data System (ADS)

Chen, Chao; Sun, Shan; Ji, Xiaoquan; Yin, Zejie

2017-08-01

A novel real time magnetic island identification system for HL-2A is introduced. The identification method is based on the measurement of Mirnov probes and the equilibrium flux constructed by the equilibrium fit (EFIT) code. The system consists of an analog front board and a digital processing board connected by a shield cable. Four octal-channel analog-to-digital convertors are utilized for 100 KHz simultaneous sampling of all the probes, and the applications of PCI extensions for Instrumentation platform and reflective memory allow the system to receive EFIT results simultaneously. A high performance field programmable gate array (FPGA) is used to realize the real time identification algorithm. Based on the parallel and pipeline processing of the FPGA, the magnetic island structure can be identified with a cycle time of 3 ms during experiments.

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.

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.

Edge Thomson scattering diagnostic on COMPASS tokamak: Installation, calibration, operation, improvements

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

Bohm, P., E-mail: bohm@ipp.cas.cz; Bilkova, P.; Melich, R.

2014-11-15

The core Thomson scattering diagnostic (TS) on the COMPASS tokamak was put in operation and reported earlier. Implementation of edge TS, with spatial resolution along the laser beam up to ∼1/100 of the tokamak minor radius, is presented now. The procedure for spatial calibration and alignment of both core and edge systems is described. Several further upgrades of the TS system, like a triggering unit and piezo motor driven vacuum window shutter, are introduced as well. The edge TS system, together with the core TS, is now in routine operation and provides electron temperature and density profiles.

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.

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.

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

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.

Development of the PARVMEC Code for Rapid Analysis of 3D MHD Equilibrium

NASA Astrophysics Data System (ADS)

Seal, Sudip; Hirshman, Steven; Cianciosa, Mark; Wingen, Andreas; Unterberg, Ezekiel; Wilcox, Robert; ORNL Collaboration

2015-11-01

The VMEC three-dimensional (3D) MHD equilibrium has been used extensively for designing stellarator experiments and analyzing experimental data in such strongly 3D systems. Recent applications of VMEC include 2D systems such as tokamaks (in particular, the D3D experiment), where application of very small (delB/B ~ 10-3) 3D resonant magnetic field perturbations render the underlying assumption of axisymmetry invalid. In order to facilitate the rapid analysis of such equilibria (for example, for reconstruction purposes), we have undertaken the task of parallelizing the VMEC code (PARVMEC) to produce a scalable and temporally rapidly convergent equilibrium code for use on parallel distributed memory platforms. The parallelization task naturally splits into three distinct parts 1) radial surfaces in the fixed-boundary part of the calculation; 2) two 2D angular meshes needed to compute the Green's function integrals over the plasma boundary for the free-boundary part of the code; and 3) block tridiagonal matrix needed to compute the full (3D) pre-conditioner near the final equilibrium state. Preliminary results show that scalability is achieved for tasks 1 and 3, with task 2 still nearing completion. The impact of this work on the rapid reconstruction of D3D plasmas using PARVMEC in the V3FIT code will be discussed. Work supported by U.S. DOE under Contract DE-AC05-00OR22725 with UT-Battelle, LLC.

Study of steam condensation at sub-atmospheric pressure: setting a basic research using MELCOR code

NASA Astrophysics Data System (ADS)

Manfredini, A.; Mazzini, M.

2017-11-01

One of the most serious accidents that can occur in the experimental nuclear fusion reactor ITER is the break of one of the headers of the refrigeration system of the first wall of the Tokamak. This results in water-steam mixture discharge in vacuum vessel (VV), with consequent pressurization of this container. To prevent the pressure in the VV exceeds 150 KPa absolute, a system discharges the steam inside a suppression pool, at an absolute pressure of 4.2 kPa. The computer codes used to analyze such incident (eg. RELAP 5 or MELCOR) are not validated experimentally for such conditions. Therefore, we planned a basic research, in order to have experimental data useful to validate the heat transfer correlations used in these codes. After a thorough literature search on this topic, ACTA, in collaboration with the staff of ITER, defined the experimental matrix and performed the design of the experimental apparatus. For the thermal-hydraulic design of the experiments, we executed a series of calculations by MELCOR. This code, however, was used in an unconventional mode, with the development of models suited respectively to low and high steam flow-rate tests. The article concludes with a discussion of the placement of experimental data within the map featuring the phenomenon characteristics, showing the importance of the new knowledge acquired, particularly in the case of chugging.

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.

The Dynamic Mutation Characteristics of Thermonuclear Reaction in Tokamak

PubMed Central

Li, Jing; Quan, Tingting; Zhang, Wei; Deng, Wei

2014-01-01

The stability and bifurcations of multiple limit cycles for the physical model of thermonuclear reaction in Tokamak are investigated in this paper. The one-dimensional Ginzburg-Landau type perturbed diffusion equations for the density of the plasma and the radial electric field near the plasma edge in Tokamak are established. First, the equations are transformed to the average equations with the method of multiple scales and the average equations turn to be a Z 2-symmetric perturbed polynomial Hamiltonian system of degree 5. Then, with the bifurcations theory and method of detection function, the qualitative behavior of the unperturbed system and the number of the limit cycles of the perturbed system for certain groups of parameter are analyzed. At last, the stability of the limit cycles is studied and the physical meaning of Tokamak equations under these parameter groups is given. PMID:24892099

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.

Resonance localization in tokamaks excited with ICRF waves

NASA Astrophysics Data System (ADS)

Kerbel, G. D.; McCoy, M. G.

1985-06-01

Advanced wave model used to evaluate ICRH in tokamaks typically used warm plasma theory and allow inhomogeneity in one dimension. The majority of these calculations neglect the fact that gyrocenters experience the inhomogeneity via their motion parallel to the magnetic field. In strongly driven systems, wave damping can distort the particle distribution function supporting the wave and this produces changes in the absorption. A bounce-averaged Fokker-Planck quasilinear computational model which evolves the population of particles on more realistic orbits is presented. Each wave-particle resonance has its own specific interaction amplitude within any given volume element; these data need only be generated once, and appropriately stored for efficient retrieval. The wave-particle resonant interaction then serves as a mechanism by which the diffusion of particle populations can proceed among neighboring orbits. The local specific spectral energy absorption rate is directly calculable once the orbit geometry and populations are determined. The code is constructed in such fashion as to accommodate wave propagation models which provide the wave spectral energy density on a poloidal cross-section. Information provided by the calculation includes the local absorption properties of the medium which can then be exploited to evolve the wave field.

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

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.

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.

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.

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

Low-Frequency Microinstabilities in Rotating Tokamak Plasmas.

NASA Astrophysics Data System (ADS)

Artun, Mehmet

1994-01-01

Low-frequency drift-type microinstabilities have often been suggested as the leading candidates to account for the anomalously large transport; observed in tokamak plasmas. The effects of sheared equilibrium flows on this important class of instabilities is systematically investigated in the present thesis. In particular, the analysis is carried out in two parts. In order to gain some insight into the key elements of this problem, the first part deals with the stability properties of the kinetic ion temperature gradient mode under the influence of parallel and perpendicular shear flows in a simplified sheared magnetic slab geometry. The eigenmode analysis is performed using a shooting code for long-wavelength modes (k_|rho _{i} << 1), and an integral eigenmode code for short-wavelength modes (k_ |rho_{i} ~ 1). Numerical results are cross-checked with analytical estimates in the fluid regime. While the differential analysis is mostly limited to ground state modes of the system--due to the requirement that the average perpendicular wavenumber be small--the integral eigenmode code has been used to calculate higher radial eigenmodes with confidence. New features observed through the introduction of shear flows are discussed. In the second part we present the shear flow generalization of the nonlinear electromagnetic gyrokinetic equation for realistic toroidal geometry. In accordance with the most natural choice for such studies, the coordinate frame is chosen to be shifted in velocity space and unchanged in configuration space. The natural equilibrium constraints of the toroidal problem limits the choice of the flow profile to that in which the angular velocity is a function of the flux surface. The general form of the gyrokinetic equation obtained is then used to derive the two-dimensional linear electrostatic eigenmode equation in circular toroidal geometry including trapped particle effects. In addition to magnetic trapping, electrostatic and centrifugal trapping are also found to play an important role here. A modified version of a finite element code is utilized to analyze shear flow effects on the trapped ion mode (TIM) in the long wavelength limit. Numerical results for fully coupled as well as single poloidal harmonic cases are presented. Implications of the results obtained in the present investigation are discussed and suggestions are given for future studies.

Sub-millisecond electron density profile measurement at the JET tokamak with the fast lithium beam emission spectroscopy system

NASA Astrophysics Data System (ADS)

Réfy, D. I.; Brix, M.; Gomes, R.; Tál, B.; Zoletnik, S.; Dunai, D.; Kocsis, G.; Kálvin, S.; Szabolics, T.; JET Contributors

2018-04-01

Diagnostic alkali atom (e.g., lithium) beams are routinely used to diagnose magnetically confined plasmas, namely, to measure the plasma electron density profile in the edge and the scrape off layer region. A light splitting optics system was installed into the observation system of the lithium beam emission spectroscopy diagnostic at the Joint European Torus (JET) tokamak, which allows simultaneous measurement of the beam light emission with a spectrometer and a fast avalanche photodiode (APD) camera. The spectrometer measurement allows density profile reconstruction with ˜10 ms time resolution, absolute position calculation from the Doppler shift, spectral background subtraction as well as relative intensity calibration of the channels for each discharge. The APD system is capable of measuring light intensities on the microsecond time scale. However ˜100 μs integration is needed to have an acceptable signal to noise ratio due to moderate light levels. Fast modulation of the beam up to 30 kHz is implemented which allows background subtraction on the 100 μs time scale. The measurement covers the 0.9 < ρpol < 1.1 range with 6-10 mm optical resolution at the measurement location which translates to 3-5 mm radial resolution at the midplane due to flux expansion. An automated routine has been developed which performs the background subtraction, the relative calibration, and the comprehensive error calculation, runs a Bayesian density reconstruction code, and loads results to the JET database. The paper demonstrates the capability of the APD system by analyzing fast phenomena like pellet injection and edge localized modes.

CompactPCI/Linux Platform in FTU Slow Control System

NASA Astrophysics Data System (ADS)

Iannone, F.; Wang, L.; Centioli, C.; Panella, M.; Mazza, G.; Vitale, V.

2004-12-01

In large fusion experiments, such as tokamak devices, there is a common trend for slow control systems. Because of complexity of the plants, the so-called `Standard Model' (SM) in slow control has been adopted on several tokamak machines. This model is based on a three-level hierarchical control: 1) High-Level Control (HLC) with a supervisory function; 2) Medium-Level Control (MLC) to interface and concentrate I/O field equipments; 3) Low-Level Control (LLC) with hard real-time I/O function, often managed by PLCs. FTU control system designed with SM concepts has underwent several stages of developments in its fifteen years duration of runs. The latest evolution was inevitable, due to the obsolescence of the MLC CPUs, based on VME-MOTOROLA 68030 with OS9 operating system. A large amount of C code was developed for that platform to route the data flow from LLC, which is constituted by 24 Westinghouse Numalogic PC-700 PLCs with about 8000 field-points, to HLC, based on a commercial Object-Oriented Real-Time database on Alpha/CompaqTru64 platform. Therefore, we have to look for cost-effective solutions and finally a CompactPCI-Intel x86 platform with Linux operating system was chosen. A software porting has been done, taking into account the differences between OS9 and Linux operating system in terms of Inter/Network Processes Communications and I/O multi-ports serial driver. This paper describes the hardware/software architecture of the new MLC system, emphasizing the reliability and the low costs of the open source solutions. Moreover, a huge amount of software packages available in open source environment will assure a less painful maintenance, and will open the way to further improvements of the system itself.

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.

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.

Design and Implementation of a 200kW, 28GHz gyrotron system for the Compact Toroidal Hybrid Experiment

NASA Astrophysics Data System (ADS)

Hartwell, G. J.; Knowlton, S. F.; Ennis, D. A.; Maurer, D. A.; Bigelow, T.

2016-10-01

The Compact Toroidal Hybrid (CTH) is an l = 2 , m = 5 torsatron/tokamak hybrid (R0 = 0.75 m, ap 0.2 m, and | B | <= 0.7 T). It can generate its highly configurable confining magnetic fields solely with external coils, but typically operates with up to 80 kA of ohmically-generated plasma current for heating. New studies of edge plasma transport in stellarator geometries will benefit from CTH operating as a pure torsatron with a high temperature edge plasma. Accordingly, a 28 GHz, 200 kW gyrotron operating at 2nd harmonic for ECRH is being installed to supplement the existing 15 kW klystron system operating at the fundamental frequency; the latter will be used to initially generate the plasma. Ray-tracing calculations that guide the selection of launching position, antenna focal length, and beam-steering characteristics of the ECRH have been performed with the TRAVIS code [ 1 ] . The calculated absorption is up to 95.7% for vertically propagating rays, however, the absorption is more sensitive to magnetic field variations than for a side launch where the field gradient is tokamak-like. The design of the waveguide path and components for the top-launch scenario will be presented. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

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.

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

Tokamak foundation in USSR/Russia 1950-1990

NASA Astrophysics Data System (ADS)

Smirnov, V. P.

2010-01-01

In the USSR, nuclear fusion research began in 1950 with the work of I.E. Tamm, A.D. Sakharov and colleagues. They formulated the principles of magnetic confinement of high temperature plasmas, that would allow the development of a thermonuclear reactor. Following this, experimental research on plasma initiation and heating in toroidal systems began in 1951 at the Kurchatov Institute. From the very first devices with vessels made of glass, porcelain or metal with insulating inserts, work progressed to the operation of the first tokamak, T-1, in 1958. More machines followed and the first international collaboration in nuclear fusion, on the T-3 tokamak, established the tokamak as a promising option for magnetic confinement. Experiments continued and specialized machines were developed to test separately improvements to the tokamak concept needed for the production of energy. At the same time, research into plasma physics and tokamak theory was being undertaken which provides the basis for modern theoretical work. Since then, the tokamak concept has been refined by a world-wide effort and today we look forward to the successful operation of ITER.

The computation in diagnostics for tokamaks: systems, designs, approaches

NASA Astrophysics Data System (ADS)

Krawczyk, Rafał; Linczuk, Paweł; Czarski, Tomasz; Wojeński, Andrzej; Chernyshova, Maryna; Poźniak, Krzysztof; Kolasiński, Piotr; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Kowalska-Strzeciwilk, Ewa; Malinowski, Karol; Gaska, Michał

2017-08-01

The requirements given for GEM (Gaseous Electron Multiplier) detector based acquisition system for plasma impurities diagnostics triggered a need for the development of a specialized software and hardware architecture. The amount of computations with latency and throughput restrictions cause that an advanced solution is sought for. In order to provide a mechanism fitting the designated tokamaks, an insight into existing solutions was necessary. In the article there is discussed architecture of systems used for plasma diagnostics and in related scientific fields. The developed solution is compared and contrasted with other diagnostic and control systems. Particular attention is payed to specific requirements for plasma impurities diagnostics in tokamak thermal fusion reactor. Subsequently, the details are presented that justified the choice of the system architecture and the discussion on various approaches is given.

Long-distance delivery of multi-channel polarization signals in nuclear fusion research

NASA Astrophysics Data System (ADS)

Ko, Jinseok; Chung, Jinil; Lee, Kyuhang

2017-04-01

A polarization-preserving optical system that includes a dual photoelastic modulator (PEM) has been designed and fabricated for the motional Stark effect (MSE) diagnostic system which measures internal magnetic field structures inside the tokamak for the Korea Superconducting Tokamak Advanced Research. The collection optics located outside the vacuum window is composed of four lenses, a dielectric coated mirror, and a dichroic beam splitter in addition to the PEM and a polarizer. The fiber dissector is designed based on the focal plane that aligns 25 lines of sight, each of which constitutes a bundle of 19 600-μm fibers. The fibers run about 40 m from the front optics in the tokamak vacuum vessel to the detector in the diagnostic area remote from the tokamak hall. This takes the advantage of the fact that the polarization information is intensity-modulated once going through the PEM and the polarizer. The polarization signals measured by the MSE diagnostic successfully demonstrates its proof-of-principle physics that is critical in the stable and steady-state operation of the tokamak plasmas.

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.

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

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

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.

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.

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.

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

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.

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

Improved operating scenarios of the DIII-D tokamak as a result of the addition of UNIX computer systems

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

Henline, P.A.

1995-12-31

The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DIII-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape controlmore » due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described.« less

Improved operating scenarios of the DIII-D tokamak as a result of the addition of UNIX computer systems

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

Henline, P.A.

1995-10-01

The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DRI-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape controlmore » due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described.« less

Status of parallel Python-based implementation of UEDGE

NASA Astrophysics Data System (ADS)

Umansky, M. V.; Pankin, A. Y.; Rognlien, T. D.; Dimits, A. M.; Friedman, A.; Joseph, I.

2017-10-01

The tokamak edge transport code UEDGE has long used the code-development and run-time framework Basis. However, with the support for Basis expected to terminate in the coming years, and with the advent of the modern numerical language Python, it has become desirable to move UEDGE to Python, to ensure its long-term viability. Our new Python-based UEDGE implementation takes advantage of the portable build system developed for FACETS. The new implementation gives access to Python's graphical libraries and numerical packages for pre- and post-processing, and support of HDF5 simplifies exchanging data. The older serial version of UEDGE has used for time-stepping the Newton-Krylov solver NKSOL. The renovated implementation uses backward Euler discretization with nonlinear solvers from PETSc, which has the promise to significantly improve the UEDGE parallel performance. We will report on assessment of some of the extended UEDGE capabilities emerging in the new implementation, and will discuss the future directions. Work performed for U.S. DOE by LLNL under contract DE-AC52-07NA27344.

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.

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).

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.

Chaotic density fluctuations in L-mode plasmas of the DIII-D tokamak

DOE PAGES

Maggs, J. E.; Rhodes, Terry L.; Morales, G. J.

2015-03-05

Analysis of the time series obtained with the Doppler backscattering system (DBS) in the DIII-D tokamak shows that intermediate wave number plasma density fluctuations in low confinement (L-mode) tokamak plasmas are chaotic. Here, the supporting evidence is based on the shape of the power spectrum; the location of the signal in the complexity-entropy plane (C-H plane); and the population of the corresponding Bandt-Pompe probability distributions.

Adaptive optimal stochastic state feedback control of resistive wall modes in tokamaks

NASA Astrophysics Data System (ADS)

Sun, Z.; Sen, A. K.; Longman, R. W.

2006-01-01

An adaptive optimal stochastic state feedback control is developed to stabilize the resistive wall mode (RWM) instability in tokamaks. The extended least-square method with exponential forgetting factor and covariance resetting is used to identify (experimentally determine) the time-varying stochastic system model. A Kalman filter is used to estimate the system states. The estimated system states are passed on to an optimal state feedback controller to construct control inputs. The Kalman filter and the optimal state feedback controller are periodically redesigned online based on the identified system model. This adaptive controller can stabilize the time-dependent RWM in a slowly evolving tokamak discharge. This is accomplished within a time delay of roughly four times the inverse of the growth rate for the time-invariant model used.

Adaptive Optimal Stochastic State Feedback Control of Resistive Wall Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Sun, Z.; Sen, A. K.; Longman, R. W.

2007-06-01

An adaptive optimal stochastic state feedback control is developed to stabilize the resistive wall mode (RWM) instability in tokamaks. The extended least square method with exponential forgetting factor and covariance resetting is used to identify the time-varying stochastic system model. A Kalman filter is used to estimate the system states. The estimated system states are passed on to an optimal state feedback controller to construct control inputs. The Kalman filter and the optimal state feedback controller are periodically redesigned online based on the identified system model. This adaptive controller can stabilize the time dependent RWM in a slowly evolving tokamak discharge. This is accomplished within a time delay of roughly four times the inverse of the growth rate for the time-invariant model used.

Pre-Results of the Real-Time ODIN Validation on MARTe Using Plasma Linearized Model in FTU Tokamak

NASA Astrophysics Data System (ADS)

Sadeghi, Yahya; Boncagni, Luca

2012-06-01

MARTe is a modular framework for real-time control aspects. At present time there are several MARTe systems under development at Frascati Tokamak Upgrade (Boncagni et al. in First steps in the FTU migration towards a modular and distributed real time control architecture based on MARTe and RTNet, 2010) such as the LH power percentage system, the gas puffing control system, the real-time ODIN plasma equilibrium reconstruction system and the position/current feedback control system (in a design phase) (Boncagni et al. in J Fusion Eng Design). The real-time reconstruction of magnetic flux in FTU tokamak is an important issue to estimate some quantities that can be use to control the plasma. This paper addresses the validation of real-time implementation of that task on MARTe.

Probing spherical tokamak plasmas using charged fusion products

NASA Astrophysics Data System (ADS)

Boeglin, Werner U.; Perez, Ramona V.; Darrow, Douglass S.; Cecconello, Marco; Klimek, Iwona; Allan, Scott Y.; Akers, Rob J.; Jones, Owen M.; Keeling, David L.; McClements, Ken G.; Scannell, Rory

2015-11-01

The detection of charged fusion products, such as protons and tritons resulting from D(d,p)t reactions, can be used to determine the fusion reaction rate profile in large spherical tokamak plasmas with neutral beam heating. The time resolution of a diagnostic of this type makes it possible to study the slowly-varying beam density profile, as well as rapid changes resulting from MHD instabilities. A 4-channel prototype proton detector (PD) was installed and operated on the MAST spherical tokamak in August/September 2013, and a new 6-channel system for the NSTX-U spherical tokamak is under construction. PD and neutron camera measurements obtained on MAST will be compared with TRANSP calculations, and the design of the new NSTX-U system will be presented, together with the first results from this diagnostic, if available. Supported in part by DOE DE-SC0001157.

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.

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.

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

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

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.

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.

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.

PARVMEC: An Efficient, Scalable Implementation of the Variational Moments Equilibrium Code

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

Seal, Sudip K; Hirshman, Steven Paul; Wingen, Andreas

The ability to sustain magnetically confined plasma in a state of stable equilibrium is crucial for optimal and cost-effective operations of fusion devices like tokamaks and stellarators. The Variational Moments Equilibrium Code (VMEC) is the de-facto serial application used by fusion scientists to compute magnetohydrodynamics (MHD) equilibria and study the physics of three dimensional plasmas in confined configurations. Modern fusion energy experiments have larger system scales with more interactive experimental workflows, both demanding faster analysis turnaround times on computational workloads that are stressing the capabilities of sequential VMEC. In this paper, we present PARVMEC, an efficient, parallel version of itsmore » sequential counterpart, capable of scaling to thousands of processors on distributed memory machines. PARVMEC is a non-linear code, with multiple numerical physics modules, each with its own computational complexity. A detailed speedup analysis supported by scaling results on 1,024 cores of a Cray XC30 supercomputer is presented. Depending on the mode of PARVMEC execution, speedup improvements of one to two orders of magnitude are reported. PARVMEC equips fusion scientists for the first time with a state-of-theart capability for rapid, high fidelity analyses of magnetically confined plasmas at unprecedented scales.« less

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.

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.

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.

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

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.

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).

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.

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

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.

Electronic system for Langmuir probe measurements

NASA Astrophysics Data System (ADS)

Mitov, M.; Bankova, A.; Dimitrova, M.; Ivanova, P.; Tutulkov, K.; Djermanova, N.; Dejarnac, R.; Stöckel, J.; Popov, Tsv K.

2012-03-01

A newly developed Langmuir probe system for measurements of current-voltage (IV) characteristics in the tokamak divertor area is presented and discussed. The system is partially controlled by a computer allowing simultaneous and independent feeding and registration of signals. The system is mounted in the COMPASS tokamak, Institute of Plasma Physics, Academy of Sciences of the Czech Republic. The new electronic circuit boards include also active low-pass filters which smooth the signal before recording by the data acquisition system (DAQ). The signal is thus less noisy and the data processing is much easier. We also designed and built a microcontroller-driven waveform generator with resolution of 1 Ms/s. The power supply is linear and uses a transformer. We avoided the use of a switching power supply because of the noise that it could generate. Examples of measurements of the IV characteristics by divertor probes in the COMPASS tokamak and evaluation of the EEDF are presented.

Effect of ECRH and resonant magnetic fields on formation of magnetic islands in the T-10 tokamak plasma

NASA Astrophysics Data System (ADS)

Shestakov, E. A.; Savrukhin, P. V.

2017-10-01

Experiments in the T-10 tokamak demonstrated possibility of controlling the plasma current during disruption instability using the electron cyclotron resonance heating (ECRH) and the controlled operation of the ohmic current-holding system. Quasistable plasma discharge with repeating sawtooth oscillations can be restored after energy quench using auxiliary ECRH power when PEC / POH > 2-5. The external magnetic field generation system consisted of eight saddle coils that were arranged symmetrically relative to the equatorial plane of the torus outside of the vacuum vessel of the T-10 tokamak to study the possible resonant magnetic field effects on the rotation frequency of magnetic islands. The saddle coils power supply system is based on four thyristor converters with a total power of 300 kW. The power supply control system is based on Siemens S7 controllers. As shown by preliminary experiments, the interaction efficiency of external magnetic fields with plasma depends on the plasma magnetic configuration. Optimal conditions for slowing the rotation of magnetic islands were determined. Additionally, the direction of the error magnetic field in the T-10 tokamak was determined, and the threshold value of the external magnetic field was determined.

Fast Time Response Electromagnetic Disruption Mitigation Concept

DOE PAGES

Raman, R.; Jarboe, T.; Jernigan, Thomas C.; ...

2015-09-28

An important and urgent issue for ITER is predicting and controlling disruptions. Tokamaks and spherical tokamaks have the potential to disrupt. Methods to rapidly quench the discharge after an impending disruption is detected are essential to protect the vessel and internal components. The warning time for the onset of some disruptions in tokamaks could be <10 ms, which poses stringent requirements on the disruption mitigation system for reactor systems. In this proposed method, a cylindrical boron nitride projectile containing a radiative payload composed of boron, boron nitride, or beryllium particulate matter and weighing similar to 15 g is accelerated tomore » velocities on the order of 1 to 2 km/s in <2 ms in a linear rail gun accelerator. A partially fragmented capsule is then injected into the tokamak discharge in the 3- to 6-ms timescale, where the radiative payload is dispersed. The device referred to as an electromagnetic particle injector has the potential to meet the short warning timescales for which a reactor disruption mitigation system must be built. The system is fully electromagnetic, with no mechanical moving parts, which ensures high reliability after a period of long standby.« 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

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

Phase 3 experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics. Volume 1: Experiment

NASA Astrophysics Data System (ADS)

Oyama, Yukio; Konno, Chikara; Ikeda, Yujiro; Maekawa, Fujio; Kosako, Kazuaki; Nakamura, Tomoo; Maekawa, Hiroshi; Youssef, Mahmoud Z.; Kumar, Anil; Abdou, Mohamed A.

1994-02-01

A pseudo-line source has been realized by using an accelerator based D-T point neutron source. The pseudo-line source is obtained by time averaging of continuously moving point source or by superposition of finely distributed point sources. The line source is utilized for fusion blanket neutronics experiments with an annular geometry so as to simulate a part of a tokamak reactor. The source neutron characteristics were measured for two operational modes for the line source, continuous and step-wide modes, with the activation foil and the NE213 detectors, respectively. In order to give a source condition for a successive calculational analysis on the annular blanket experiment, the neutron source characteristics was calculated by a Monte Carlo code. The reliability of the Monte Carlo calculation was confirmed by comparison with the measured source characteristics. The shape of the annular blanket system was a rectangular with an inner cavity. The annular blanket was consist of 15 mm-thick first wall (SS304) and 406 mm-thick breeder zone with Li2O at inside and Li2CO3 at outside. The line source was produced at the center of the inner cavity by moving the annular blanket system in the span of 2 m. Three annular blanket configurations were examined; the reference blanket, the blanket covered with 25 mm thick graphite armor and the armor-blanket with a large opening. The neutronics parameters of tritium production rate, neutron spectrum and activation reaction rate were measured with specially developed techniques such as multi-detector data acquisition system, spectrum weighting function method and ramp controlled high voltage system. The present experiment provides unique data for a higher step of benchmark to test a reliability of neutronics design calculation for a realistic tokamak reactor.

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.

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.

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.

ICANT, a code for the self-consistent computation of ICRH antenna coupling

NASA Astrophysics Data System (ADS)

Pécoul, S.; Heuraux, S.; Koch, R.; Leclert, G.

1996-02-01

The code deals with 3D antenna structures (finite length antennae) that are used to launch electromagnetic waves into tokamak plasmas. The antenna radiation problem is solved using a finite boundary element technique combined with a spectral solution of the interior problem. The slab approximation is used, and periodicity in y and z directions is introduced to account for toroidal geometry. We present results for various types of antennae radiating in vacuum: antenna with a finite Faraday screen and ideal Faraday screen, antenna with side limiters and phased antenna arrays. The results (radiated power, current profile) obtained are very close to analytical solutions when available.

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.

A predictive transport modeling code for ICRF-heated tokamaks

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

Phillips, C.K.; Hwang, D.Q.; Houlberg, W.

In this report, a detailed description of the physic included in the WHIST/RAZE package as well as a few illustrative examples of the capabilities of the package will be presented. An in depth analysis of ICRF heating experiments using WHIST/RAZE will be discussed in a forthcoming report. A general overview of philosophy behind the structure of the WHIST/RAZE package, a summary of the features of the WHIST code, and a description of the interface to the RAZE subroutines are presented in section 2 of this report. Details of the physics contained in the RAZE code are examined in section 3.more » Sample results from the package follow in section 4, with concluding remarks and a discussion of possible improvements to the package discussed in section 5.« less

A predictive transport modeling code for ICRF-heated tokamaks

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

Phillips, C.K.; Hwang, D.Q.; Houlberg, W.

1992-02-01

In this report, a detailed description of the physic included in the WHIST/RAZE package as well as a few illustrative examples of the capabilities of the package will be presented. An in depth analysis of ICRF heating experiments using WHIST/RAZE will be discussed in a forthcoming report. A general overview of philosophy behind the structure of the WHIST/RAZE package, a summary of the features of the WHIST code, and a description of the interface to the RAZE subroutines are presented in section 2 of this report. Details of the physics contained in the RAZE code are examined in section 3.more » Sample results from the package follow in section 4, with concluding remarks and a discussion of possible improvements to the package discussed in section 5.« less

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.

Modeling RF Fields in Hot Plasmas with Parallel Full Wave Code

NASA Astrophysics Data System (ADS)

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

2016-10-01

FAR-TECH, Inc. is developing a suite of full wave RF plasma codes. It is based on a meshless formulation in configuration space with adapted cloud of computational points (CCP) capability and using the hot plasma conductivity kernel to model the nonlocal plasma dielectric response. The conductivity kernel is calculated by numerically integrating the linearized Vlasov equation along unperturbed particle trajectories. Work has been done on the following calculations: 1) the conductivity kernel in hot plasmas, 2) a monitor function based on analytic solutions of the cold-plasma dispersion relation, 3) an adaptive CCP based on the monitor function, 4) stencils to approximate the wave equations on the CCP, 5) the solution to the full wave equations in the cold-plasma model in tokamak geometry for ECRH and ICRH range of frequencies, and 6) the solution to the wave equations using the calculated hot plasma conductivity kernel. We will present results on using a meshless formulation on adaptive CCP to solve the wave equations and on implementing the non-local hot plasma dielectric response to the wave equations. The presentation will include numerical results of wave propagation and absorption in the cold and hot tokamak plasma RF models, using DIII-D geometry and plasma parameters. Work is supported by the U.S. DOE SBIR program.

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

TFTR CAMAC systems and components

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

Rauch, W.A.; Bergin, W.; Sichta, P.

1987-08-01

Princeton's tokamak fusion test reactor (TFTR) utilizes Computer Automated Measurement and Control (CAMAC) to provide instrumentation for real and quasi real time control, monitoring, and data acquisition systems. This paper describes and discusses the complement of CAMAC hardware systems and components that comprise the interface for tokamak control and measurement instrumentation, and communication with the central instrumentation control and data acquisition (CICADA) system. It also discusses CAMAC reliability and calibration, types of modules used, a summary of data acquisition and control points, and various diagnostic maintenance tools used to support and troubleshoot typical CAMAC systems on TFTR.

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.

Modeling of toroidal torques exerted by internal kink instability in a tokamak plasma

NASA Astrophysics Data System (ADS)

Zhang, N.; Liu, Y. Q.; Yu, D. L.; Wang, S.; Xia, G. L.; Dong, G. Q.; Bai, X.

2017-08-01

Toroidal modeling efforts are initiated to systematically compute and compare various toroidal torques, exerted by an unstable internal kink in a tokamak plasma, using the MARS-F/K/Q suite of codes. The torques considered here include the resonant electromagnetic torque due to the Maxwell stress (the EM or JXB torque), the neoclassical toroidal viscous (NTV) torque, and the torque associated with the Reynolds stress. Numerical results show that the relative magnitude of the net resonant electromagnetic and the Reynolds stress torques increases with the equilibrium flow speed of the plasma, whilst the net NTV torque follows the opposite trend. The global flow shear sensitively affects the Reynolds stress torque, but not the electromagnetic and the NTV torques. Detailed examinations reveal dominant contributions to the Maxwell and Reynolds stress torques, in terms of the poloidal harmonic numbers of various perturbation fields, as well as their relative toroidal phasing.

Evidence of coupling to Global Alfv{acute e}ne Eigenmodes during Alfv{acute e}n wave current drive experiments on the Phaedrus-T tokamak

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

Vukovic, M.; Wukitch, S.; Harper, M.

1996-02-01

A series of experiments designed to explore mechanisms of power deposition during Alfv{acute e}n wave current drive experiments on the Phaedrus-T tokamak has shown evidence of power deposition via mode conversion of Global Alfv{acute e}n Eigenmodes at the Alfv{acute e}n resonance. Observation of radially localized RF induced density fluctuations in the plasma and their location vs. {ital B}{sub {ital T}} is in agreement with the predictions of behaviour of GAE damping on the AR by the toroidal code LION. Furthermore, the change in the time evolution of the loop voltage, is consistent with the change of effective power deposition radius,more » {ital r}{sub PD}, and is in agreement with the density fluctuations radius. {copyright} {ital 1996 American Institute of Physics.}« 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.

Kinetic equilibrium reconstruction for the NBI- and ICRH-heated H-mode plasma on EAST tokamak

NASA Astrophysics Data System (ADS)

Zhen, ZHENG; Nong, XIANG; Jiale, CHEN; Siye, DING; Hongfei, DU; Guoqiang, LI; Yifeng, WANG; Haiqing, LIU; Yingying, LI; Bo, LYU; Qing, ZANG

2018-04-01

The equilibrium reconstruction is important to study the tokamak plasma physical processes. To analyze the contribution of fast ions to the equilibrium, the kinetic equilibria at two time-slices in a typical H-mode discharge with different auxiliary heatings are reconstructed by using magnetic diagnostics, kinetic diagnostics and TRANSP code. It is found that the fast-ion pressure might be up to one-third of the plasma pressure and the contribution is mainly in the core plasma due to the neutral beam injection power is primarily deposited in the core region. The fast-ion current contributes mainly in the core region while contributes little to the pedestal current. A steep pressure gradient in the pedestal is observed which gives rise to a strong edge current. It is proved that the fast ion effects cannot be ignored and should be considered in the future study of EAST.

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.

Ion cyclotron emission from energetic fusion products in tokamak plasmas: A full-wave calculation

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

Batchelor, D.B.; Jaeger, E.F.; Colestock, P.L.

1989-06-01

A full-wave ion cyclotron resonant heating (ICRH) code has been modified to allow calculation of cyclotron emission from energetic ions in tokamaks. The immediate application is to fusion alpha particles in near-ignition devices. This permits detailed evaluation of proposed alpha particle diagnostics (Proceedings of the Thirteenth European Conference on Controlled Fusion and Plasma Heating, Schliersee, Federal Republic of Germany, 1986, edited by G. Briffod and M. Kaufmann (European Physical Society, Petit-Lancy, Switzerland, 1986), Part 1, Vol. 2, p. 37.) This full-wave approach automatically takes into account wall reflections, standing waves, and plasma absorption and overcomes the difficulties inherent in attemptingmore » to apply conventional geometrical optics to long wavelengths. By calculating the coherent radiation field caused by an ensemble of localized current sources (and retaining the phase information), the directivity of pickup antennas is correctly represented.« less

Potential Application of a Graphical Processing Unit to Parallel Computations in the NUBEAM Code

NASA Astrophysics Data System (ADS)

Payne, J.; McCune, D.; Prater, R.

2010-11-01

NUBEAM is a comprehensive computational Monte Carlo based model for neutral beam injection (NBI) in tokamaks. NUBEAM computes NBI-relevant profiles in tokamak plasmas by tracking the deposition and the slowing of fast ions. At the core of NUBEAM are vector calculations used to track fast ions. These calculations have recently been parallelized to run on MPI clusters. However, cost and interlink bandwidth limit the ability to fully parallelize NUBEAM on an MPI cluster. Recent implementation of double precision capabilities for Graphical Processing Units (GPUs) presents a cost effective and high performance alternative or complement to MPI computation. Commercially available graphics cards can achieve up to 672 GFLOPS double precision and can handle hundreds of thousands of threads. The ability to execute at least one thread per particle simultaneously could significantly reduce the execution time and the statistical noise of NUBEAM. Progress on implementation on a GPU will be presented.

Damping Rate Measurements of Medium n Alfv'en Eigenmodes in JET

NASA Astrophysics Data System (ADS)

Klein, Alexander; Testa, Duccio; Snipes, Joseph; Fasoli, Ambrogio; Carfantan, Hervé

2007-11-01

Alfv'en Eigenmodes (AE's) with mode numbers 5 < n < 20 are expected to be unstable in burning tokamaks and may lead to loss of fast particle confinement. The active MHD spectroscopy program at JET has already provided a wealth of information about low n (n <= 2) AE's in the past decade, but a recently installed array of four antennas is capable of driving higher mode numbered (n < 100, 30 < f < 350 kHz) perturbations. In the latest JET campaign, the damping rates for several types of AE's were measured parasitically in a wide range of tokamak scenarios. We review the active MHD diagnostic and present the first measurements of medium-n AE stability on JET, then describe future plans for the active MHD spectroscopy project. The data analysis involves a novel method for resolving multiple AE's that exist at identical frequencies, which uses techniques based on the SparSpec code.

The development of a universal diagnostic probe system for Tokamak fusion test reactor

NASA Technical Reports Server (NTRS)

Mastronardi, R.; Cabral, R.; Manos, D.

1982-01-01

The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

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

SUNIST Microwave Power System

NASA Astrophysics Data System (ADS)

Feng, Songlin; Yang, Xuanzong; Feng, Chunhua; Wang, Long; Rao, Jun; Feng, Kecheng

2005-06-01

Experiments on the start-up and formation of spherical tokamak plasmas by electron cyclotron heating alone without ohmic heating and electrode discharge assisted electron cyclotron wave current start-up will be carried out on the SUNIST (Sino United Spherical Tokamak) device. The 2.45 GHz/100kW/30 ms microwave power system and 1000 V/50 A power supply for electrode discharge are ready for experiments with non-inductive current drive.

Gamma ray imager on the DIII-D tokamak

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

Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.

2016-04-15

A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electronsmore » in the energy range of 1–60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.« less

Gamma ray imager on the DIII-D tokamak

DOE PAGES

Pace, D. C.; Cooper, C. M.; Taussig, D.; ...

2016-04-13

A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electronsmore » in the energy range of 1- 60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. In conclusion, first measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.« less

Stainless steel blanket concept for tokamaks

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

Karbowski, J.S.; Lee, A.Y.; Prevenslik, T.V.

1979-01-25

The purpose of this joint ORNL/Westinghouse Program is to develop a design concept for a tokamak reactor blanket system which satisfies engineering requirements for a utility environment. While previous blanket studies have focused primarily on performance issues (thermal, neutronic, and structural), this study has emphasized consideration of reliability, fabricability, and lifetime.

Helical core reconstruction of a DIII-D hybrid scenario tokamak discharge

DOE PAGES

Cianciosa, Mark; Wingen, Andreas; Hirshman, Steven P.; ...

2017-05-18

Our paper presents the first fully 3-dimensional (3D) equilibrium reconstruction of a helical core in a tokamak device. Using a new parallel implementation of the Variational Moments Equilibrium Code (PARVMEC) coupled to V3FIT, 3D reconstructions can be performed at resolutions necessary to produce helical states in nominally axisymmetric tokamak equilibria. In a flux pumping experiment performed on DIII-D, an external n=1 field was applied while a 3/2 neoclassical tearing mode was suppressed using ECCD. The externally applied field was rotated past a set of fixed diagnostics at a 20 Hz frequency. Furthermore, the modulation, were found to be strongest in the core SXR and MSE channels, indicates a localized rotating 3D structure locked in phase with the applied field. Signals from multiple time slices are converted to a virtual rotation of modeled diagnostics adding 3D signal information. In starting from an axisymmetric equilibrium reconstruction solution, the reconstructed broader current profile flattens the q-profile, resulting in an m=1, n=1 perturbation of the magnetic axis that ismore » $$\\sim 50\\times $$ larger than the applied n=1 deformation of the edge. Error propagation confirms that the displacement of the axis is much larger than the uncertainty in the axis position validating the helical equilibrium.« less

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.

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.

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.

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.

Helical core reconstruction of a DIII-D hybrid scenario tokamak discharge

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

Cianciosa, Mark; Wingen, Andreas; Hirshman, Steven P.

Our paper presents the first fully 3-dimensional (3D) equilibrium reconstruction of a helical core in a tokamak device. Using a new parallel implementation of the Variational Moments Equilibrium Code (PARVMEC) coupled to V3FIT, 3D reconstructions can be performed at resolutions necessary to produce helical states in nominally axisymmetric tokamak equilibria. In a flux pumping experiment performed on DIII-D, an external n=1 field was applied while a 3/2 neoclassical tearing mode was suppressed using ECCD. The externally applied field was rotated past a set of fixed diagnostics at a 20 Hz frequency. Furthermore, the modulation, were found to be strongest in the core SXR and MSE channels, indicates a localized rotating 3D structure locked in phase with the applied field. Signals from multiple time slices are converted to a virtual rotation of modeled diagnostics adding 3D signal information. In starting from an axisymmetric equilibrium reconstruction solution, the reconstructed broader current profile flattens the q-profile, resulting in an m=1, n=1 perturbation of the magnetic axis that ismore » $$\\sim 50\\times $$ larger than the applied n=1 deformation of the edge. Error propagation confirms that the displacement of the axis is much larger than the uncertainty in the axis position validating the helical equilibrium.« less

Calculations of Alfven Wave Driving Forces, Plasma Flow and Current Drive in Tokamak Plasmas

NASA Astrophysics Data System (ADS)

Elfimov, Artur; Galvao, Ricardo; Amarante-Segundo, Gesil; Nascimento, Ivan

2000-10-01

A general form of time-averaged poloidal ponderomotive forces induced by fast and kinetic Alfvin waves by direct numerical calculations and in geometric optics approximation are analyzed on the basis of the collisionless two fluid (ions and electrons) magneto-hydrodynamics equation. Analytical approximations are used to clarify the effect of Larmour radius on radio-frequency (RF) ponderomotive forces and on poloidal flows induced by them in tokamak plasmas.The RF ponderomotive force is expressed as a sum of a gradient part and of a wave momentum transfer force, which is proportional to wave dissipation. The gradient electromagnetic stress force is combined with fluid dynamic (Reynolds) stress force. It is shown that accounting only Reynolds stress term can overestimate the plasma flow and it is found that the finite ion Larmor radius effect play fundamental role in ponderomotive forces that can drive a poloidal flow, which is larger than a flow driven by a wave momentum transfer force. Finally, balancing the RF forces by the electron-ion friction and viscous force the current and plasma flows driven by ponderomotive forces are calculated for tokamak plasmas, using a kinetic code [Phys. Plasmas, v.6 (1999) p.2437]. Strongly sheared current and plasma flow waves is found.

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.

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

Suppression of runaway electrons with a resonant magnetic perturbation in MST tokamak plasmas

NASA Astrophysics Data System (ADS)

Munaretto, Stefano; Chapman, B. E.; Almagri, A. F.; Cornille, B. S.; Dubois, A. M.; Goetz, J. A.; McCollam, K. J.; Sovinec, C. R.

2016-10-01

Runaway electrons generated in MST tokamak plasmas are now being probed with resonant magnetic perturbations (RMP's). An RMP with m =3 strongly suppresses the runaway electrons. Initial modeling of these plasmas with NIMROD shows the degradation of flux surfaces with an m =3 RMP, which may account for the runaway electron suppression. These MST tokamak plasmas have Bt =0.14 T, Ip =50kA, and q(a) =2.2, with a bulk electron density and temperature of 5x1017 m-3 and 150 eV. Runaway electrons are detected via x-ray emission. The RMP is produced by a poloidal array of 32 saddle coils at the narrow vertical insulated cut in MST's thick conducting shell. Each RMP has a single m but a broad n spectrum. A sufficiently strong m =3 RMP completely suppresses the runaway electrons, while a comparable m =1 RMP has little effect. The impact of the RMP's on the magnetic topology of these plasmas is being studied with the nonlinear MHD code, NIMROD. With an m =3 RMP, stochasticity is introduced in the outer third of the plasma. No such change is observed with the m =1 RMP. NIMROD also predicts regularly occurring sawtooth oscillations with a period comparable to MHD activity observed in the experiment. Work supported by USDOE.

Structure of chaotic magnetic field lines in IR-T1 tokamak due to ergodic magnetic limiter

NASA Astrophysics Data System (ADS)

Ahmadi, S.; Salar Elahi, A.; Ghorannevis, M.

2018-03-01

In this paper we have studied an Ergodic Magnetic Limiter (EML) based chaotic magnetic field for transport control in the edge plasma of IR-T1 tokamak. The resonance created by the EML causes perturbation of the equilibrium field line in tokamak and as a result, the field lines are chaotic in the vicinity of the dimerized island chains. Transport barriers are formed in the chaotic field line and actually observe in tokamak with reverse magnetic shear. We used area-preserving non-twist (and twist) Poincaré maps to describe the formation of transport barriers, which are actually features of Hamiltonian systems. This transport barrier is useful in reducing radial diffusion of the field line and thus improving the plasma confinement.

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).

Coupling Effect between Equilibrium Field and Heating Field and Modification of the Power Supply System on SUNIST Spherical Tokamak

NASA Astrophysics Data System (ADS)

He, Yexi; Li, Xiaoyan; Gao, Zhe

2005-02-01

Strong inductive coupling between the heating field and equilibrium field is confirmed to be responsible for the poor plasma equilibrium in initial discharges on the SUNIST spherical tokamak. A modification project for the power supply system of equilibrium field coils is successfully performed to increase the duration time of plasma current flattop from much less than 1ms to about 2 ms.

Phase Contrast Imaging on the HL-2A Tokamak

NASA Astrophysics Data System (ADS)

Yu, Yi; Gong, Shaobo; Xu, Min; Jiang, Wei; Zhong, Wulv; Shi, Zhongbin; Wang, Huajie; Wu, Yifan; Yuan, Boda; Lan, Tao; Ye, Minyou; Duan, Xuru; HL-2A Team

2016-10-01

In this article we present the design of a phase contrast imaging (PCI) system on the HL-2A tokamak. This diagnostic is developed to infer line integrated plasma density fluctuations by measuring the phase shift of an expanded CO2 laser beam passing through magnetically confined high temperature plasmas. This system is designed to diagnose plasma density fluctuations with the maximum wavenumber of 66 cm-1. The designed wavenumber resolution is 2.09cm-1, and the time resolution is higher than 0.2 μs. The broad kρs ranging from 0.34 to 13.37 makes it suitable for turbulence measurement. An upgraded PCI system is also discussed, which is designed for the HL-2M tokamak. Supported by the National Magnetic Confinement Fusion Energy Research Project (Grant No. 2015GB120002), the National Natural Science Foundation of China (Grant No. 11375053, 11105144, 10905057, 11535013).

Energy, Vacuum, Gas Fueling, and Security Systems for the Spherical Tokamak MEDUSA-CR

NASA Astrophysics Data System (ADS)

Gonzalez, Jeferson; Soto, Christian; Carvajal, Johan; Ribeiro, Celso

2013-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5 T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We present here the energy, vacuum, gas fueling, and security systems for MEDUSA-CR device. The interface with the control and data acquisition systems based on National Instruments (NI) software (LabView) and hardware (on loan to our laboratory via NI-Costa Rica) are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

Alternate fusion fuels workshop

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

Not Available

1981-06-01

The workshop was organized to focus on a specific confinement scheme: the tokamak. The workshop was divided into two parts: systems and physics. The topics discussed in the systems session were narrowly focused on systems and engineering considerations in the tokamak geometry. The workshop participants reviewed the status of system studies, trade-offs between d-t and d-d based reactors and engineering problems associated with the design of a high-temperature, high-field reactor utilizing advanced fuels. In the physics session issues were discussed dealing with high-beta stability, synchrotron losses and transport in alternate fuel systems. The agenda for the workshop is attached.

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.

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.

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.

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

Fast island phase identification for tearing mode feedback control on J-TEXT tokamak

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

Rao, B., E-mail: borao@hust.edu.cn; Li, D.; Hu, F. R.

A new method to control the tearing mode (TM) in tokamaks has been proposed [Q. Hu and Q. Yu, Nucl. Fusion 56, 034001 (5pp.) (2016)], according to which, the external resonant magnetic perturbation needs to be applied in certain magnetic island phase regions. Therefore, it is very important to identify the helical phase of magnetic islands in real time. The TM in tokamak plasmas is normally rotating and carries magnetic oscillations, which are known as Mirnov oscillations and can be detected by Mirnov probes. When the O-point or X-point of the magnetic island passes through the probe, the signal willmore » experience a zero-crossing. A poloidal Mirnov probe array and a corresponding island phase identification method are presented. A field-programmable gate array is used to provide the magnetic island helical phase in real time by using multichannel zero crossing detection. This system has been developed on the J-TEXT tokamak and works well. This paper introduces the establishment of the fast magnetic island phase identifying system.« less

Control system of neoclassical tearing modes in real time on HL-2A tokamak.

PubMed

Yan, Longwen; Ji, Xiaoquan; Song, Shaodong; Xia, Fan; Xu, Yuan; Ye, Jiruo; Jiang, Min; Chen, Wenjin; Sun, Tengfei; Liang, Shaoyong; Ling, Fei; Ma, Rui; Huang, Mei; Qu, Hongpeng; Song, Xianming; Yu, Deliang; Shi, Zhongbin; Liu, Yi; Yang, Qingwei; Xu, Min; Duan, Xuru; Liu, Yong

2017-11-01

The stability and performance of tokamak plasmas are routinely limited by various magneto-hydrodynamic instabilities, such as neoclassical tearing modes (NTMs). This paper presents a rather simple method to control the NTMs in real time (RT) on a tokamak, including the control principle of a feedback approach for RT suppression and stabilization for the NTMs. The control system combines Mirnov, electron cyclotron emission, and soft X-ray diagnostics used for determining the NTM positions. A methodology for fast detection of 2/1 or 3/2 NTM positions with 129 × 129 grid reconstruction is elucidated. The forty poloidal angles for steering the electron cyclotron resonance heating (ECRH)/electron cyclotron current drive launcher are used to establish the alignment of antenna mirrors with the center of the NTM and to ensure launcher emission intersecting with the rational surface of a magnetic island. Pilot experiments demonstrate the RT control capability to trace the conventional tearing modes (CTMs) in the HL-2A tokamak. The 2/1 CTMs have been suppressed or stabilized by the ECRH power deposition on site or with the steerable launcher.

Control system of neoclassical tearing modes in real time on HL-2A tokamak

NASA Astrophysics Data System (ADS)

Yan, Longwen; Ji, Xiaoquan; Song, Shaodong; Xia, Fan; Xu, Yuan; Ye, Jiruo; Jiang, Min; Chen, Wenjin; Sun, Tengfei; Liang, Shaoyong; Ling, Fei; Ma, Rui; Huang, Mei; Qu, Hongpeng; Song, Xianming; Yu, Deliang; Shi, Zhongbin; Liu, Yi; Yang, Qingwei; Xu, Min; Duan, Xuru; Liu, Yong

2017-11-01

The stability and performance of tokamak plasmas are routinely limited by various magneto-hydrodynamic instabilities, such as neoclassical tearing modes (NTMs). This paper presents a rather simple method to control the NTMs in real time (RT) on a tokamak, including the control principle of a feedback approach for RT suppression and stabilization for the NTMs. The control system combines Mirnov, electron cyclotron emission, and soft X-ray diagnostics used for determining the NTM positions. A methodology for fast detection of 2/1 or 3/2 NTM positions with 129 × 129 grid reconstruction is elucidated. The forty poloidal angles for steering the electron cyclotron resonance heating (ECRH)/electron cyclotron current drive launcher are used to establish the alignment of antenna mirrors with the center of the NTM and to ensure launcher emission intersecting with the rational surface of a magnetic island. Pilot experiments demonstrate the RT control capability to trace the conventional tearing modes (CTMs) in the HL-2A tokamak. The 2/1 CTMs have been suppressed or stabilized by the ECRH power deposition on site or with the steerable launcher.

Design and optimization of Artificial Neural Networks for the modelling of superconducting magnets operation in tokamak fusion reactors

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

Froio, A.; Bonifetto, R.; Carli, S.

In superconducting tokamaks, the cryoplant provides the helium needed to cool different clients, among which by far the most important one is the superconducting magnet system. The evaluation of the transient heat load from the magnets to the cryoplant is fundamental for the design of the latter and the assessment of suitable strategies to smooth the heat load pulses, induced by the intrinsically pulsed plasma scenarios characteristic of today's tokamaks, is crucial for both suitable sizing and stable operation of the cryoplant. For that evaluation, accurate but expensive system-level models, as implemented in e.g. the validated state-of-the-art 4C code, weremore » developed in the past, including both the magnets and the respective external cryogenic cooling circuits. Here we show how these models can be successfully substituted with cheaper ones, where the magnets are described by suitably trained Artificial Neural Networks (ANNs) for the evaluation of the heat load to the cryoplant. First, two simplified thermal-hydraulic models for an ITER Toroidal Field (TF) magnet and for the ITER Central Solenoid (CS) are developed, based on ANNs, and a detailed analysis of the chosen networks' topology and parameters is presented and discussed. The ANNs are then inserted into the 4C model of the ITER TF and CS cooling circuits, which also includes active controls to achieve a smoothing of the variation of the heat load to the cryoplant. The training of the ANNs is achieved using the results of full 4C simulations (including detailed models of the magnets) for conventional sigmoid-like waveforms of the drivers and the predictive capabilities of the ANN-based models in the case of actual ITER operating scenarios are demonstrated by comparison with the results of full 4C runs, both with and without active smoothing, in terms of both accuracy and computational time. Exploiting the low computational effort requested by the ANN-based models, a demonstrative optimization study has been finally carried out, with the aim of choosing among different smoothing strategies for the standard ITER plasma operation.« less

Neoclassical toroidal viscosity in perturbed equilibria with general tokamak geometry

NASA Astrophysics Data System (ADS)

Logan, Nikolas C.; Park, Jong-Kyu; Kim, Kimin; Wang, Zhirui; Berkery, John W.

2013-12-01

This paper presents a calculation of neoclassical toroidal viscous torque independent of large-aspect-ratio expansions across kinetic regimes. The Perturbed Equilibrium Nonambipolar Transport (PENT) code was developed for this purpose, and is compared to previous combined regime models as well as regime specific limits and a drift kinetic δf guiding center code. It is shown that retaining general expressions, without circular large-aspect-ratio or other orbit approximations, can be important at experimentally relevant aspect ratio and shaping. The superbanana plateau, a kinetic resonance effect recently recognized for its relevance to ITER, is recovered by the PENT calculations and shown to require highly accurate treatment of geometric effects.

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.

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.

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

Kim, Dong-Hwan; Hong, Suk-Ho; National Fusion Research Institute

Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliabilitymore » of the method.« less

A distributed control system for the lower-hybrid current drive system on the Tokamak de Varennes

NASA Astrophysics Data System (ADS)

Bagdoo, J.; Guay, J. M.; Chaudron, G.-A.; Decoste, R.; Demers, Y.; Hubbard, A.

1990-08-01

An rf current drive system with an output power of 1 MW at 3.7 GHz is under development for the Tokamak de Varennes. The control system is based on an Ethernet local-area network of programmable logic controllers as front end, personal computers as consoles, and CAMAC-based DSP processors. The DSP processors ensure the PID control of the phase and rf power of each klystron, and the fast protection of high-power rf hardware, all within a 40 μs loop. Slower control and protection, event sequencing and the run-time database are provided by the programmable logic controllers, which communicate, via the LAN, with the consoles. The latter run a commercial process-control console software. The LAN protocol respects the first four layers of the ISO/OSI 802.3 standard. Synchronization with the tokamak control system is provided by commercially available CAMAC timing modules which trigger shot-related events and reference waveform generators. A detailed description of each subsystem and a performance evaluation of the system will be presented.

Solenoid-free plasma start-up in spherical tokamaks

NASA Astrophysics Data System (ADS)

Raman, R.; Shevchenko, V. F.

2014-10-01

The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

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

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-19

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritiummore » allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.« 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.

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

Lucia, M., E-mail: mlucia@pppl.gov; Kaita, R.; Majeski, R.

The Materials Analysis and Particle Probe (MAPP) is a compact in vacuo surface science diagnostic, designed to provide in situ surface characterization of plasma facing components in a tokamak environment. MAPP has been implemented for operation on the Lithium Tokamak Experiment at Princeton Plasma Physics Laboratory (PPPL), where all control and analysis systems are currently under development for full remote operation. Control systems include vacuum management, instrument power, and translational/rotational probe drive. Analysis systems include onboard Langmuir probes and all components required for x-ray photoelectron spectroscopy, low-energy ion scattering spectroscopy, direct recoil spectroscopy, and thermal desorption spectroscopy surface analysis techniques.

Transition to subcritical turbulence in a tokamak plasma

NASA Astrophysics Data System (ADS)

van Wyk, F.; Highcock, E. G.; Schekochihin, A. A.; Roach, C. M.; Field, A. R.; Dorland, W.

2016-12-01

Tokamak turbulence, driven by the ion-temperature gradient and occurring in the presence of flow shear, is investigated by means of local, ion-scale, electrostatic gyrokinetic simulations (with both kinetic ions and electrons) of the conditions in the outer core of the Mega-Ampere Spherical Tokamak (MAST). A parameter scan in the local values of the ion-temperature gradient and flow shear is performed. It is demonstrated that the experimentally observed state is near the stability threshold and that this stability threshold is nonlinear: sheared turbulence is subcritical, i.e. the system is formally stable to small perturbations, but, given a large enough initial perturbation, it transitions to a turbulent state. A scenario for such a transition is proposed and supported by numerical results: close to threshold, the nonlinear saturated state and the associated anomalous heat transport are dominated by long-lived coherent structures, which drift across the domain, have finite amplitudes, but are not volume filling; as the system is taken away from the threshold into the more unstable regime, the number of these structures increases until they overlap and a more conventional chaotic state emerges. Whereas this appears to represent a new scenario for transition to turbulence in tokamak plasmas, it is reminiscent of the behaviour of other subcritically turbulent systems, e.g. pipe flows and Keplerian magnetorotational accretion flows.

Extracting 3D Information from 1D and 2D Diagnostic Systems on the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Brookman, Michael

2017-10-01

The interpretation of tokamak data often hinges on assumptions of axisymetry and flux surface equilibria, neglecting 3D effects. This work discusses examples on the DIII-D tokamak where this assumption is an insufficient approximation, and explores the diagnostic information available to resolve 3D effects while preserving 1D profiles. Methods for extracting 3D data from the electron cyclotron emission radiometers, density profile reflectometer, and Thomson scattering system are discussed. Coordinating diagnostics around the tokamak shows the significance of 3D features, such as sawteeth[1] and resonant magnetic perturbations. A consequence of imposed 3D perturbations is a shift in major radius of measured profiles between diagnostics at different toroidal locations. Integrating different diagnostics requires a database containing information about their toroidal, poloidal, and radial locations. Through the data analysis framework OMFIT, it is possible to measure the magnitude of the apparent shifts from 3D effects and enforce consistency between diagnostics. Using the existing 1D and 2D diagnostic systems on DIII-D, this process allows the effects of the 3D perturbations on 1D profiles to be addressed. Supported by US DOE contracts DE-FC02-04ER54698, DE-FG03-97ER54415.

Coupling between a multi-physics workflow engine and an optimization framework

NASA Astrophysics Data System (ADS)

Di Gallo, L.; Reux, C.; Imbeaux, F.; Artaud, J.-F.; Owsiak, M.; Saoutic, B.; Aiello, G.; Bernardi, P.; Ciraolo, G.; Bucalossi, J.; Duchateau, J.-L.; Fausser, C.; Galassi, D.; Hertout, P.; Jaboulay, J.-C.; Li-Puma, A.; Zani, L.

2016-03-01

A generic coupling method between a multi-physics workflow engine and an optimization framework is presented in this paper. The coupling architecture has been developed in order to preserve the integrity of the two frameworks. The objective is to provide the possibility to replace a framework, a workflow or an optimizer by another one without changing the whole coupling procedure or modifying the main content in each framework. The coupling is achieved by using a socket-based communication library for exchanging data between the two frameworks. Among a number of algorithms provided by optimization frameworks, Genetic Algorithms (GAs) have demonstrated their efficiency on single and multiple criteria optimization. Additionally to their robustness, GAs can handle non-valid data which may appear during the optimization. Consequently GAs work on most general cases. A parallelized framework has been developed to reduce the time spent for optimizations and evaluation of large samples. A test has shown a good scaling efficiency of this parallelized framework. This coupling method has been applied to the case of SYCOMORE (SYstem COde for MOdeling tokamak REactor) which is a system code developed in form of a modular workflow for designing magnetic fusion reactors. The coupling of SYCOMORE with the optimization platform URANIE enables design optimization along various figures of merit and constraints.

Conductor analysis of the ITER FEAT poloidal field coils during a plasma scenario

NASA Astrophysics Data System (ADS)

Nicollet, S.; Hertout, P.; Duchateau, J. L.; Bleyer, A.; Bessette, D.

2002-05-01

In the framework of the ITER (International Thermonuclear Experimental Reactor) FEAT (Fusion Energy Advanced Tokamak) project, a fully superconducting PF (Poloidal Field) system has been designed in detail. The Central Solenoid and the 6 equilibrium coils constituting the PF system provide the magnetic fields which develop, shape and control the 15 MA plasma during the 1800 s of a typical plasma scenario. The 6 PF coils will be wound two-in-hand from a 45 kA niobium-titanium CICC (Cable-In-Conduit-Conductor). These coils will experience severe heat loads specially during the 400 s of the plasma burn: nuclear heating due to the 400 MW of fusion power, thermal radiation and AC losses (30 to 300 kJ). The AC losses along the PF coil pancakes are deduced from accurate magnetic field computations performed with a 3D magnetostatic code, TRAPS. The nuclear heating and the thermal radiation are assumed to be uniform over a given face of the PF coils. These heat loads are used as input to perform the thermal and hydraulic analysis with a finite element code, GANDALF. The temperature increases (0.1 to 0.4 K) are computed, the margins and performances of the conductor are evaluated.

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

Alfvén eigenmode evolution computed with the VENUS and KINX codes for the ITER baseline scenario

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

Isaev, M. Yu., E-mail: isaev-my@nrcki.ru; Medvedev, S. Yu.; Cooper, W. A.

A new application of the VENUS code is described, which computes alpha particle orbits in the perturbed electromagnetic fields and its resonant interaction with the toroidal Alfvén eigenmodes (TAEs) for the ITER device. The ITER baseline scenario with Q = 10 and the plasma toroidal current of 15 MA is considered as the most important and relevant for the International Tokamak Physics Activity group on energetic particles (ITPA-EP). For this scenario, typical unstable TAE-modes with the toroidal index n = 20 have been predicted that are localized in the plasma core near the surface with safety factor q = 1.more » The spatial structure of ballooning and antiballooning modes has been computed with the ideal MHD code KINX. The linear growth rates and the saturation levels taking into account the damping effects and the different mode frequencies have been calculated with the VENUS code for both ballooning and antiballooning TAE-modes.« less

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

A comparison of data interoperability approaches of fusion codes with application to synthetic diagnostics

NASA Astrophysics Data System (ADS)

Kruger, Scott; Shasharina, S.; Vadlamani, S.; McCune, D.; Holland, C.; Jenkins, T. G.; Candy, J.; Cary, J. R.; Hakim, A.; Miah, M.; Pletzer, A.

2010-11-01

As various efforts to integrate fusion codes proceed worldwide, standards for sharing data have emerged. In the U.S., the SWIM project has pioneered the development of the Plasma State, which has a flat-hierarchy and is dominated by its use within 1.5D transport codes. The European Integrated Tokamak Modeling effort has developed a more ambitious data interoperability effort organized around the concept of Consistent Physical Objects (CPOs). CPOs have deep hierarchies as needed by an effort that seeks to encompass all of fusion computing. Here, we discuss ideas for implementing data interoperability that is complementary to both the Plasma State and CPOs. By making use of attributes within the netcdf and HDF5 binary file formats, the goals of data interoperability can be achieved with a more informal approach. In addition, a file can be simultaneously interoperable to several standards at once. As an illustration of this approach, we discuss its application to the development of synthetic diagnostics that can be used for multiple codes.

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

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

The radiation asymmetry in MGI rapid shutdown on J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Tong, Ruihai; Chen, Zhongyong; Huang, Duwei; Cheng, Zhifeng; Zhang, Xiaolong; Zhuang, Ge; J-TEXT Team

2017-10-01

Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER. The mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, the radiated energy during the thermal quench (TQ) in massive gas injection (MGI) induced disruptions is found toroidal asymmetric, and the degrees of asymmetry correlate with the gas penetration and MGI induced magnetohydrodynamics (MHD) activities. A toroidal and poloidal array of ultraviolet photodiodes (AXUV) has been developed to investigate the radiation asymmetry on J-TEXT tokamak. Together with the upgraded mirnov probe arrays, the relation between MGI triggered MHD activities with radiation asymmetry is studied.

Development of full wave code for modeling RF fields in hot non-uniform plasmas

NASA Astrophysics Data System (ADS)

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

2016-10-01

FAR-TECH, Inc. is developing a full wave RF modeling code to model RF fields in fusion devices and in general plasma applications. As an important component of the code, an adaptive meshless technique is introduced to solve the wave equations, which allows resolving plasma resonances efficiently and adapting to the complexity of antenna geometry and device boundary. The computational points are generated using either a point elimination method or a force balancing method based on the monitor function, which is calculated by solving the cold plasma dispersion equation locally. Another part of the code is the conductivity kernel calculation, used for modeling the nonlocal hot plasma dielectric response. The conductivity kernel is calculated on a coarse grid of test points and then interpolated linearly onto the computational points. All the components of the code are parallelized using MPI and OpenMP libraries to optimize the execution speed and memory. The algorithm and the results of our numerical approach to solving 2-D wave equations in a tokamak geometry will be presented. Work is supported by the U.S. DOE SBIR program.

Full Wave Parallel Code for Modeling RF Fields in Hot Plasmas

NASA Astrophysics Data System (ADS)

Spencer, Joseph; Svidzinski, Vladimir; Evstatiev, Evstati; Galkin, Sergei; Kim, Jin-Soo

2015-11-01

FAR-TECH, Inc. is developing a suite of full wave RF codes in hot plasmas. It is based on a formulation in configuration space with grid adaptation capability. The conductivity kernel (which includes a nonlocal dielectric response) is calculated by integrating the linearized Vlasov equation along unperturbed test particle orbits. For Tokamak applications a 2-D version of the code is being developed. Progress of this work will be reported. This suite of codes has the following advantages over existing spectral codes: 1) It utilizes the localized nature of plasma dielectric response to the RF field and calculates this response numerically without approximations. 2) It uses an adaptive grid to better resolve resonances in plasma and antenna structures. 3) It uses an efficient sparse matrix solver to solve the formulated linear equations. The linear 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. Work is supported by the U.S. DOE SBIR program.

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.

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.

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.

Chaotic coordinates for the Large Helical Device

NASA Astrophysics Data System (ADS)

Hudson, Stuart; Suzuki, Yasuhiro

2014-10-01

The study of dynamical systems is facilitated by a coordinate framework with coordinate surfaces that coincide with invariant structures of the dynamical flow. For axisymmetric systems, a continuous family of invariant surfaces is guaranteed and straight-fieldline coordinates may be constructed. For non-integrable systems, e.g. stellarators, perturbed tokamaks, this continuous family is broken. Nevertheless, coordinates can still be constructed that simplify the description of the dynamics. The Poincare-Birkhoff theorem, the Aubry-Mather theorem, and the KAM theorem show that there are important structures that are invariant under the perturbed dynamics; namely the periodic orbits, the cantori, and the irrational flux surfaces. Coordinates adapted to these invariant sets, which we call chaotic coordinates, provide substantial advantages. The regular motion becomes straight, and the irregular motion is bounded by, and dissected by, coordinate surfaces that coincide with surfaces of locally-minimal magnetic-fieldline flux. The chaotic edge of the magnetic field, as calculated by HINT2 code, in the Large Helical Device (LHD) is examined, and a coordinate system is constructed so that the flux surfaces are ``straight'' and the islands become ``square.''

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

How much does a tokamak reactor cost?

NASA Astrophysics Data System (ADS)

Freidberg, J.; Cerfon, A.; Ballinger, S.; Barber, J.; Dogra, A.; McCarthy, W.; Milanese, L.; Mouratidis, T.; Redman, W.; Sandberg, A.; Segal, D.; Simpson, R.; Sorensen, C.; Zhou, M.

2017-10-01

The cost of a fusion reactor is of critical importance to its ultimate acceptability as a commercial source of electricity. While there are general rules of thumb for scaling both overnight cost and levelized cost of electricity the corresponding relations are not very accurate or universally agreed upon. We have carried out a series of scaling studies of tokamak reactor costs based on reasonably sophisticated plasma and engineering models. The analysis is largely analytic, requiring only a simple numerical code, thus allowing a very large number of designs. Importantly, the studies are aimed at plasma physicists rather than fusion engineers. The goals are to assess the pros and cons of steady state burning plasma experiments and reactors. One specific set of results discusses the benefits of higher magnetic fields, now possible because of the recent development of high T rare earth superconductors (REBCO); with this goal in mind, we calculate quantitative expressions, including both scaling and multiplicative constants, for cost and major radius as a function of central magnetic field.

Hydrocarbon deposition in gaps of tungsten and graphite tiles in Experimental Advanced Superconducting Tokamak edge plasma parameters

NASA Astrophysics Data System (ADS)

Xu, Qian; Yang, Zhongshi; Luo, Guang-Nan

2015-09-01

The three-dimensional (3D) Monte Carlo code PIC-EDDY has been utilized to investigate the mechanism of hydrocarbon deposition in gaps of tungsten tiles in the Experimental Advanced Superconducting Tokamak (EAST), where the sheath potential is calculated by the 2D in space and 3D in velocity particle-in-cell method. The calculated results for graphite tiles using the same method are also presented for comparison. Calculation results show that the amount of carbon deposited in the gaps of carbon tiles is three times larger than that in the gaps of tungsten tiles when the carbon particles from re-erosion on the top surface of monoblocks are taken into account. However, the deposition amount is found to be larger in the gaps of tungsten tiles at the same CH4 flux. When chemical sputtering becomes significant as carbon coverage on tungsten increases with exposure time, the deposition inside the gaps of tungsten tiles would be considerable.

Design of the Cryostat for HT-7U Superconducting Tokamak

NASA Astrophysics Data System (ADS)

Yu, Jie; Wu, Song-tao; Song, Yun-tao; Weng, Pei-de

2002-06-01

The cryostat of HT-7U tokamak is a large vacuum vessel surrounding the entire basic machine with a cylindrical shell, a dished top and a flat bottom. The main function of HT-7U cryostat is to provide a thermal barrier between an ambient temperature test hall and a liquid helium-cooled superconducting magnet. The loads applied to the cryostat are from sources of vacuum pressure, dead weight, seismic events and electromagnetic forces originated by eddy currents. It also provides feed-through penetrations for all the connecting elements inside and outside the cryostat. The main material selected for the cryostat is stainless steel 304L. The structural analyses including buckling for the cryostat vessel under the plasma operation condition have been carried out by using a finite element code. Stress analysis results show that the maximum stress intensity was below the allowable value. In this paper, the structural analyses and design of HT-7U cryostat are emphasized.

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.

Observation of quasi-coherent edge fluctuations in Ohmic plasmas on National Spherical Torus Experiment

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

Banerjee, Santanu; Diallo, A.; Zweben, S. J.

A quasi-coherent edge density mode with frequency f{sub mode} ∼ 40 kHz is observed in Ohmic plasmas in National Spherical Torus Experiment using the gas puff imaging diagnostic. This mode is located predominantly just inside the separatrix, with a maximum fluctuation amplitude significantly higher than that of the broadband turbulence in the same frequency range. The quasi-coherent mode has a poloidal wavelength λ{sub pol} ∼ 16 cm and a poloidal phase velocity of V{sub pol} ∼ 4.9 ± 0.3 km s{sup −1} in the electron diamagnetic direction, which are similar to the characteristics expected from a linear drift-wave-like mode in the edge. This is the first observation of amore » quasi-coherent edge mode in an Ohmic diverted tokamak, and so may be useful for validating tokamak edge turbulence codes.« less

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

TRAMP; The next generation data acquisition for RTP

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

van Haren, P.C.; Wijnoltz, F.

1992-04-01

The Rijnhuizen Tokamak Project RTP is a medium-sized tokamak experiment, which requires a very reliable data-acquisition system, due to its pulsed nature. Analyzing the limitations of an existing CAMAC-based data-acquisition system showed, that substantial increase of performance and flexibility could best be obtained by the construction of an entirely new system. This paper discusses this system, CALLED TRAMP (Transient Recorder and Amoeba Multi Processor), based on tailor-made transient recorders with a multiprocessor computer system in VME running Amoeba. The performance of TRAMP exceeds the performance of the CAMAC system by a factor of four. The plans to increase the flexibilitymore » and for a further increase of performance are presented.« less

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

ICANT, a code for the self-consistent computation of ICRH antenna coupling

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

Pecoul, S.; Heuraux, S.; Koch, R.

1996-02-01

The code deals with 3D antenna structures (finite length antennae) that are used to launch electromagnetic waves into tokamak plasmas. The antenna radiation problem is solved using a finite boundary element technique combined with a spectral solution of the interior problem. The slab approximation is used, and periodicity in {ital y} and {ital z} directions is introduced to account for toroidal geometry. We present results for various types of antennae radiating in vacuum: antenna with a finite Faraday screen and ideal Faraday screen, antenna with side limiters and phased antenna arrays. The results (radiated power, current profile) obtained are verymore » close to analytical solutions when available. {copyright} {ital 1996 American Institute of Physics.}« 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.

Breakdown assisted by a novel electron drift injection in the J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Wang, Nengchao; Jin, Hai; Zhuang, Ge; Ding, Yonghua; Pan, Yuan; Cen, Yishun; Chen, Zhipeng; Huang, Hai; Liu, Dequan; Rao, Bo; Zhang, Ming; Zou, Bichen

2014-07-01

A novel electron drift injection (EDI) system aiming to improve breakdown behavior has been designed and constructed on the Joint Texas EXperiment Tokamak Tokamak. Electrons emitted by the system undergo the E×B drift, ∇B drift and curvature drift in sequence in order to traverse the confining magnetic field. A local electrostatic well, generated by a concave-shaped plate biased more negative than the cathode, is introduced to interrupt the emitted electrons moving along the magnetic field line (in the parallel direction) in an attempt to bring an enhancement of the injection efficiency and depth. A series of experiments have demonstrated the feasibility of this method, and a penetration distance deeper than 9.5 cm is achieved. Notable breakdown improvements, including the reduction of breakdown delay and average loop voltage, are observed for discharges assisted by EDI. The lower limit of successfully ionized pressure is expanded.

Application of automatic gain control for radiometer diagnostic in SST-1 tokamak.

PubMed

Makwana, Foram R; Siju, Varsha; Edappala, Praveenlal; Pathak, S K

2017-12-01

This paper describes the characterisation of a negative feedback type of automatic gain control (AGC) circuit that will be an integral part of the heterodyne radiometer system operating at a frequency range of 75-86 GHz at SST-1 tokamak. The developed AGC circuit is a combination of variable gain amplifier and log amplifier which provides both gain and attenuation typically up to 15 dB and 45 dB, respectively, at a fixed set point voltage and it has been explored for the first time in tokamak radiometry application. The other important characteristics are that it exhibits a very fast response time of 390 ns to understand the fast dynamics of electron cyclotron emission and can operate at very wide input RF power dynamic range of around 60 dB that ensures signal level within the dynamic range of the detection system.

Control and Data Acquisition for the Spherical Tokamak MEDUSA-CR

NASA Astrophysics Data System (ADS)

Soto, Christian; Gonzalez, Jeferson; Carvajal, Johan; Ribeiro, Celso

2013-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5 T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We present here the control and data acquisition systems for MEDUSA-CR device which are based on National Instruments (NI) software (LabView) and hardware on loan to our laboratory via NI-Costa Rica. The interface with the energy, gas fueling, and security systems are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

Stationary Flowing Liquid Lithium (SFLiLi) systems for tokamaks

NASA Astrophysics Data System (ADS)

Zakharov, Leonid; Gentile, Charles; Roquemore, Lane

2013-10-01

The present approach to magnetic fusion which relies on high recycling plasma-wall interaction has exhausted itself at the level of TFTR, JET, JT-60 devices with no realistic path to the burning plasma. Instead, magnetic fusion needs a return to its original idea of insulation of the plasma from the wall, which was the dominant approach in the 1970s and upon implementations has a clear path to the DEMO device with PDT ~= 100 MW and Qelectric > 1 . The SFLiLi systems of this talk is the technology tool for implementation of the guiding idea of magnetic fusion. It utilizes the unique properties of flowing LiLi to pump plasma particles and, thus, insulate plasma from the walls. The necessary flow rate, ~= 1 g3/s, is very small, thus, making the use of lithium practical and consistent with safety requirements. The talk describes how chemical activity of LiLi, which is the major technology challenge of using LiLi in tokamaks, is addressed by SFLiLi systems at the level of already performed (HT-7) experiment, and in ongoing implementations for a prototype of SFLiLi for tokamak divertors and the mid-plane limiter for EAST tokamak (to be tested in the next experimental campaign). This work is supported by US DoE contract No. DE-AC02-09-CH11466.

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

Scime, Earl E.

The magnitude and spatial dependence of neutral density in magnetic confinement fusion experiments is a key physical parameter, particularly in the plasma edge. Modeling codes require precise measurements of the neutral density to calculate charge-exchange power losses and drag forces on rotating plasmas. However, direct measurements of the neutral density are problematic. In this work, we proposed to construct a laser-based diagnostic capable of providing spatially resolved measurements of the neutral density in the edge of plasma in the DIII-D tokamak. The diagnostic concept is based on two-photon absorption laser induced fluorescence (TALIF). By injecting two beams of 205 nmmore » light (co or counter propagating), ground state hydrogen (or deuterium or tritium) can be excited from the n = 1 level to the n = 3 level at the location where the two beams intersect. Individually, the beams experience no absorption, and therefore have no difficulty penetrating even dense plasmas. After excitation, a fraction of the hydrogen atoms decay from the n = 3 level to the n = 2 level and emit photons at 656 nm (the H α line). Calculations based on the results of previous TALIF experiments in magnetic fusion devices indicated that a laser pulse energy of approximately 3 mJ delivered in 5 ns would provide sufficient signal-to-noise for detection of the fluorescence. In collaboration with the DIII-D engineering staff and experts in plasma edge diagnostics for DIII-D from Oak Ridge National Laboratory (ORNL), WVU researchers designed a TALIF system capable of providing spatially resolved measurements of neutral deuterium densities in the DIII-D edge plasma. The laser systems were specified, purchased, and assembled at WVU. The TALIF system was tested on a low-power hydrogen discharge at WVU and the plan was to move the instrument to DIII-D for installation in collaboration with ORNL researchers. After budget cuts at DIII-D, the DIII-D facility declined to support installation on their tokamak. Instead, after a no-cost extension, the apparatus was moved to the University of Washington-Seattle and successfully tested on the HIT-SI3 spheromak experiment. As a result of this project, TALIF measurements of the absolutely calibrated neutral density hydrogen and deuterium were obtained in a helicon source and in a spheromak, designs were developed for installation of a TALIF system on a tokamak, and a new, xenon-based calibration scheme was proposed and demonstrated. The xenon-calibration scheme eliminates significant problems that were identified with the standard krypton calibration scheme.« less

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.

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)

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

H-mode achievement and edge features in RFX-mod tokamak operation

NASA Astrophysics Data System (ADS)

Spolaore, M.; Cavazzana, R.; Marrelli, L.; Carraro, L.; Franz, P.; Spagnolo, S.; Zaniol, B.; Zuin, M.; Cordaro, L.; Dal Bello, S.; De Masi, G.; Ferro, A.; Finotti, C.; Grando, L.; Grenfell, G.; Innocente, P.; Kudlacek, O.; Marchiori, G.; Martines, E.; Momo, B.; Paccagnella, R.; Piovesan, P.; Piron, C.; Puiatti, M. E.; Recchia, M.; Scarin, P.; Taliercio, C.; Vianello, N.; Zanotto, L.

2017-11-01

The RFX-mod experiment is a fusion device designed to operate as a reversed field pinch (RFP), with a major radius R = 2 m and a minor radius a = 0.459 m. Its high versatility recently allowed operating it also as an ohmic tokamak, allowing comparative studies between the two configurations in the same device. The device is equipped with a state of the art MHD mode feedback control system providing a magnetic boundary effective control, by applying resonant or non-resonant magnetic perturbations (MP), both in RFP and in tokamak configurations. In the fusion community the application of MPs is widely studied as a promising tool to limit the impact of plasma filaments and ELMs (edge localized modes) on plasma facing components. An important new research line is the exploitation of the RFX-mod active control system for ELM mitigation studies. As a first step in this direction, this paper presents the most recent achievements in term of RFX-mod tokamak explored scenarios, which allowed the first investigation of the ohmic and edge biasing induced H-mode. The production of D-shaped tokamak discharges and the design and deployment of an insertable polarized electrode were accomplished. Reproducible H-mode phases were obtained with insertable electrode negative biasing in single null discharges, representing an unexplored scenario with this technique. Important modifications of the edge plasma density and flow properties are observed. During the achieved H-mode ELM-like electromagnetic composite filamentary structures are observed. They are characterized by clear vorticity and parallel current density patterns.

Development of DEMO-FNS tokamak for fusion and hybrid technologies

NASA Astrophysics Data System (ADS)

Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

2015-07-01

The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

DOE PAGES

Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; ...

2015-05-15

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10 x compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid,more » exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Finally, Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.« less

Optical layout and mechanical structure of polarimeter-interferometer system for Experimental Advanced Superconducting Tokamak

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

Zou, Z. Y.; Liu, H. Q., E-mail: hqliu@ipp.ac.cn; Jie, Y. X.

A Far-InfaRed (FIR) three-wave POlarimeter-INTerferometer (POINT) system for measurement current density profile and electron density profile is under development for the EAST tokamak. The FIR beams are transmitted from the laser room to the optical tower adjacent to EAST via ∼20 m overmoded dielectric waveguide and then divided into 5 horizontal chords. The optical arrangement was designed using ZEMAX, which provides information on the beam spot size and energy distribution throughout the optical system. ZEMAX calculations used to optimize the optical layout design are combined with the mechanical design from CATIA, providing a 3D visualization of the entire POINT system.

Optical layout and mechanical structure of polarimeter-interferometer system for Experimental Advanced Superconducting Tokamak.

PubMed

Zou, Z Y; Liu, H Q; Jie, Y X; Ding, W X; Brower, D L; Wang, Z X; Shen, J S; An, Z H; Yang, Y; Zeng, L; Wei, X C; Li, G S; Zhu, X; Lan, T

2014-11-01

A Far-InfaRed (FIR) three-wave POlarimeter-INTerferometer (POINT) system for measurement current density profile and electron density profile is under development for the EAST tokamak. The FIR beams are transmitted from the laser room to the optical tower adjacent to EAST via ∼20 m overmoded dielectric waveguide and then divided into 5 horizontal chords. The optical arrangement was designed using ZEMAX, which provides information on the beam spot size and energy distribution throughout the optical system. ZEMAX calculations used to optimize the optical layout design are combined with the mechanical design from CATIA, providing a 3D visualization of the entire POINT system.

Development of a cross-polarization scattering system for the measurement of internal magnetic fluctuations in the DIII-D tokamak

DOE PAGES

Rhodes, Terry L.; Peebles, William A.; Crocker, Neal A.; ...

2014-08-05

The design and performance of a new cross-polarization scattering (CPS) system for the localized measurement of internal magnetic fluctuations is presented. CPS is a process whereby magnetic fluctuations scatter incident electromagnetic radiation into a perpendicular polarization which is subsequently detected. A new CPS design that incorporates a unique scattering geometry was laboratory tested, optimized, and installed on the DIII-D tokamak. Plasma tests of signal-to-noise, polarization purity, and frequency response indicate proper functioning of the system. Lastly, CPS data show interesting features related to internal MHD perturbations known as sawteeth that are not observed on density fluctuations.

Development progress of the Materials Analysis and Particle Probe

NASA Astrophysics Data System (ADS)

Lucia, M.; Kaita, R.; Majeski, R.; Bedoya, F.; Allain, J. P.; Boyle, D. P.; Schmitt, J. C.; Onge, D. A. St.

2014-11-01

The Materials Analysis and Particle Probe (MAPP) is a compact in vacuo surface science diagnostic, designed to provide in situ surface characterization of plasma facing components in a tokamak environment. MAPP has been implemented for operation on the Lithium Tokamak Experiment at Princeton Plasma Physics Laboratory (PPPL), where all control and analysis systems are currently under development for full remote operation. Control systems include vacuum management, instrument power, and translational/rotational probe drive. Analysis systems include onboard Langmuir probes and all components required for x-ray photoelectron spectroscopy, low-energy ion scattering spectroscopy, direct recoil spectroscopy, and thermal desorption spectroscopy surface analysis techniques.

Development progress of the Materials Analysis and Particle Probe.

PubMed

Lucia, M; Kaita, R; Majeski, R; Bedoya, F; Allain, J P; Boyle, D P; Schmitt, J C; Onge, D A St

2014-11-01

The Materials Analysis and Particle Probe (MAPP) is a compact in vacuo surface science diagnostic, designed to provide in situ surface characterization of plasma facing components in a tokamak environment. MAPP has been implemented for operation on the Lithium Tokamak Experiment at Princeton Plasma Physics Laboratory (PPPL), where all control and analysis systems are currently under development for full remote operation. Control systems include vacuum management, instrument power, and translational/rotational probe drive. Analysis systems include onboard Langmuir probes and all components required for x-ray photoelectron spectroscopy, low-energy ion scattering spectroscopy, direct recoil spectroscopy, and thermal desorption spectroscopy surface analysis techniques.

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.

Design of geometric phase measurement in EAST Tokamak

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

Lan, T.; Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031; Liu, H. Q., E-mail: hqliu@ipp.ac.cn

2016-07-15

The optimum scheme for geometric phase measurement in EAST Tokamak is proposed in this paper. The theoretical values of geometric phase for the probe beams of EAST Polarimeter-Interferometer (POINT) system are calculated by path integration in parameter space. Meanwhile, the influences of some controllable parameters on geometric phase are evaluated. The feasibility and challenge of distinguishing geometric effect in the POINT signal are also assessed in detail.

Hybrid neural network for density limit disruption prediction and avoidance on J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Zheng, W.; Hu, F. R.; Zhang, M.; Chen, Z. Y.; Zhao, X. Q.; Wang, X. L.; Shi, P.; Zhang, X. L.; Zhang, X. Q.; Zhou, Y. N.; Wei, Y. N.; Pan, Y.; J-TEXT team

2018-05-01

Increasing the plasma density is one of the key methods in achieving an efficient fusion reaction. High-density operation is one of the hot topics in tokamak plasmas. Density limit disruptions remain an important issue for safe operation. An effective density limit disruption prediction and avoidance system is the key to avoid density limit disruptions for long pulse steady state operations. An artificial neural network has been developed for the prediction of density limit disruptions on the J-TEXT tokamak. The neural network has been improved from a simple multi-layer design to a hybrid two-stage structure. The first stage is a custom network which uses time series diagnostics as inputs to predict plasma density, and the second stage is a three-layer feedforward neural network to predict the probability of density limit disruptions. It is found that hybrid neural network structure, combined with radiation profile information as an input can significantly improve the prediction performance, especially the average warning time ({{T}warn} ). In particular, the {{T}warn} is eight times better than that in previous work (Wang et al 2016 Plasma Phys. Control. Fusion 58 055014) (from 5 ms to 40 ms). The success rate for density limit disruptive shots is above 90%, while, the false alarm rate for other shots is below 10%. Based on the density limit disruption prediction system and the real-time density feedback control system, the on-line density limit disruption avoidance system has been implemented on the J-TEXT tokamak.

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.

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.

A quasilinear operator retaining magnetic drift effects in tokamak geometry

NASA Astrophysics Data System (ADS)

Catto, Peter J.; Lee, Jungpyo; Ram, Abhay K.

2017-12-01

The interaction of radio frequency waves with charged particles in a magnetized plasma is usually described by the quasilinear operator that was originally formulated by Kennel & Engelmann (Phys. Fluids, vol. 9, 1966, pp. 2377-2388). In their formulation the plasma is assumed to be homogenous and embedded in a uniform magnetic field. In tokamak plasmas the Kennel-Engelmann operator does not capture the magnetic drifts of the particles that are inherent to the non-uniform magnetic field. To overcome this deficiency a combined drift and gyrokinetic derivation is employed to derive the quasilinear operator for radio frequency heating and current drive in a tokamak with magnetic drifts retained. The derivation requires retaining the magnetic moment to higher order in both the unperturbed and perturbed kinetic equations. The formal prescription for determining the perturbed distribution function then follows a novel procedure in which two non-resonant terms must be evaluated explicitly. The systematic analysis leads to a diffusion equation that is compact and completely expressed in terms of the drift kinetic variables. The equation is not transit averaged, and satisfies the entropy principle, while retaining the full poloidal angle variation without resorting to Fourier decomposition. As the diffusion equation is in physical variables, it can be implemented in any computational code. In the Kennel-Engelmann formalism, the wave-particle resonant delta function is either for the Landau resonance or the Doppler shifted cyclotron resonance. In the combined gyro and drift kinetic approach, a term related to the magnetic drift modifies the resonance condition.

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

The external kink mode in diverted tokamaks

DOE PAGES

Turnbull, Alan D.; Hanson, Jeremy M.; Turco, Francesca; ...

2016-06-16

Here, an explanation is provided for the disruptive instability in diverted tokamaks when the safety factor at the 95% poloidal flux surface, q 95, is driven below 2.0. The instability is a resistive kink counterpart to the current-driven ideal mode that traditionally explained the corresponding disruption in limited cross-sections when q edge, the safety factor at the outermost closed flux surface, lies just below a rational value. Experimentally, external kink modes are observed in limiter configurations as the current in a tokamak is ramped up and q edge decreases through successive rational surfaces. For q edge < 2, the instabilitymore » is always encountered and is highly disruptive. However, diverted plasmas, in which q edge is formally infinite in the magnetohydrodynamic (MHD) model, have presented a longstanding difficulty since the theory would predict stability, yet, the disruptive limit occurs in practice when q 95, reaches 2. It is shown from numerical calculations that a resistive kink mode is linearly destabilized by the rapidly increasing resistivity at the plasma edge when q 95 < 2, but q edge >> 2. 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 q = 2 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.« less

Migration of tungsten dust in tokamaks: role of dust-wall collisions

NASA Astrophysics Data System (ADS)

Ratynskaia, S.; Vignitchouk, L.; Tolias, P.; Bykov, I.; Bergsåker, H.; Litnovsky, A.; den Harder, N.; Lazzaro, E.

2013-12-01

The modelling of a controlled tungsten dust injection experiment in TEXTOR by the dust dynamics code MIGRAINe is reported. The code, in addition to the standard dust-plasma interaction processes, also encompasses major mechanical aspects of dust-surface collisions. The use of analytical expressions for the restitution coefficients as functions of the dust radius and impact velocity allows us to account for the sticking and rebound phenomena that define which parts of the dust size distribution can migrate efficiently. The experiment provided unambiguous evidence of long-distance dust migration; artificially introduced tungsten dust particles were collected 120° toroidally away from the injection point, but also a selectivity in the permissible size of transported grains was observed. The main experimental results are reproduced by modelling.

Performance and data analysis aspects of the new DIII-D monostatic profile reflectometer system

DOE PAGES

Zeng, Lei; Peebles, William A.; Doyle, Edward J.; ...

2014-08-07

A new frequency-modulated (FMCW) profile reflectometer system, featuring a monostatic antenna geometry (using one microwave antenna for both launch and receive), has been installed on the DIII-D tokamak, providing a first experimental test of this measurement approach for profile reflectometry. Significant features of the new system are briefly described in this paper, including the new monostatic arrangement, use of overmoded, broadband transmission waveguide, and dual-polarization combination/demultiplexing. Updated data processing and analysis, and in-service performance aspects of the new monostatic profile reflectometer system are also presented. By using a raytracing code (GENRAY) to determine the approximate trajectory of the probe beam,more » the electron density (n e) profile can be successfully reconstructed with L-mode plasmas vertically shifted by more than 10 cm off the vessel midplane. Specifically, it is demonstrated that the new system has a capability to measure n e profiles with plasma vertical offsets of up to ±17 cm. Furthermore, examples are also presented of accurate, high time and spatial resolution density profile measurements made over a wide range of DIII-D conditions, e.g. the measured temporal evolution of the density profile across an L-H transition.« less

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.

Cryogenic pellet launcher adapted for controlling of tokamak plasma edge instabilities.

PubMed

Lang, P T; Cierpka, P; Harhausen, J; Neuhauser, J; Wittmann, C; Gál, K; Kálvin, S; Kocsis, G; Sárközi, J; Szepesi, T; Dorner, C; Kauke, G

2007-02-01

One of the main challenges posed recently on pellet launcher systems in fusion-oriented plasma physics is the control of the plasma edge region. Strong energy bursts ejected from the plasma due to edge localized modes (ELMs) can form a severe threat for in-vessel components but can be mitigated by sufficiently frequent triggering of the underlying instabilities using hydrogen isotope pellet injection. However, pellet injection systems developed mainly for the task of ELM control, keeping the unwanted pellet fueling minimized, are still missing. Here, we report on a novel system developed under the premise of its suitability for control and mitigation of plasma edge instabilities. The system is based on the blower gun principle and is capable of combining high repetition rates up to 143 Hz with low pellet velocities. Thus, the flexibility of the accessible injection geometry can be maximized and the pellet size kept low. As a result the new system allows for an enhancement in the tokamak operation as well as for more sophisticated experiments investigating the underlying physics of the plasma edge instabilities. This article reports on the design of the new system, its main operational characteristics as determined in extensive test bed runs, and also its first test at the tokamak experiment ASDEX Upgrade.

Study of runaway electrons using the conditional average sampling method in the Damavand tokamak

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

Pourshahab, B., E-mail: bpourshahab@gmail.com; Sadighzadeh, A.; Abdi, M. R., E-mail: r.abdi@phys.ui.ac.ir

2017-03-15

Some experiments for studying the runaway electron (RE) effects have been performed using the poloidal magnetic probes system installed around the plasma column in the Damavand tokamak. In these experiments, the so-called runaway-dominated discharges were considered in which the main part of the plasma current is carried by REs. The induced magnetic effects on the poloidal pickup coils signals are observed simultaneously with the Parail–Pogutse instability moments for REs and hard X-ray bursts. The output signals of all diagnostic systems enter the data acquisition system with 2 Msample/(s channel) sampling rate. The temporal evolution of the diagnostic signals is analyzedmore » by the conditional average sampling (CAS) technique. The CASed profiles indicate RE collisions with the high-field-side plasma facing components at the instability moments. The investigation has been carried out for two discharge modes—low-toroidal-field (LTF) and high-toroidal-field (HTF) ones—related to both up and down limits of the toroidal magnetic field in the Damavand tokamak and their comparison has shown that the RE confinement is better in HTF discharges.« less

Design of set-point weighting PIλ + Dμ controller for vertical magnetic flux controller in Damavand tokamak.

PubMed

Rasouli, H; Fatehi, A

2014-12-01

In this paper, a simple method is presented for tuning weighted PI(λ) + D(μ) controller parameters based on the pole placement controller of pseudo-second-order fractional systems. One of the advantages of this controller is capability of reducing the disturbance effects and improving response to input, simultaneously. In the following sections, the performance of this controller is evaluated experimentally to control the vertical magnetic flux in Damavand tokamak. For this work, at first a fractional order model is identified using output-error technique in time domain. For various practical experiments, having desired time responses for magnetic flux in Damavand tokamak, is vital. To approach this, at first the desired closed loop reference models are obtained based on generalized characteristic ratio assignment method in fractional order systems. After that, for the identified model, a set-point weighting PI(λ) + D(μ) controller is designed and simulated. Finally, this controller is implemented on digital signal processor control system of the plant to fast/slow control of magnetic flux. The practical results show appropriate performance of this controller.

Li Experiments at the Tokamak T-11M Toward PFC Concept of Steady State Tokamak-Reactor

NASA Astrophysics Data System (ADS)

Mirnov, S. V.

2009-11-01

As practical method of using a liquid lithium as a renewable plasma-facing component (PCF) for steady state tokamak-reactor the concept of lithium emitter-collector is considered [1]. It is based on lithium filled capillary porous system proposed by V.A. Evtikhin et al. (1996). The lithium circulation process consists of four steps: (1) Li emission from the PFC emitter into the plasma; (2) plasma boundary cooling by non-coronal Li radiation; (3) Li ion capture by the collector (before they are lost to the tokamak chamber wall); (4) Li return from the collector to the emitter. T-11M tokamak experiments have used three local rail limiters made from lithium, molybdenum and graphite as lithium collectors. The lithium behavior was studied by analysis of the witness samples, and by a mobile graphite probe. The key findings are: (1) lithium collection on the ion side of the lithium limiter is 2-3 times larger than on the electron side; (2) total efficiency of Li collection integrated over all three rail limiters can reach 50-70% of the lithium emission during the discharge pulse, while the theoretical limit is about 90%. [1] S.V. Mirnov, J. Nucl. Mat., 390-391, 876 (2009).

The measurement of the intrinsic impurities of molybdenum and carbon in the Alcator C-Mod tokamak plasma using low resolution spectroscopy

NASA Astrophysics Data System (ADS)

May, M. J.; Finkenthal, M.; Regan, S. P.; Moos, H. W.; Terry, J. L.; Goetz, J. A.; Graf, M. A.; Rice, J. E.; Marmar, E. S.; Fournier, K. B.; Goldstein, W. H.

1997-06-01

The intrinsic impurity content of molybdenum and carbon was measured in the Alcator C-Mod tokamak using low resolution, multilayer mirror (MLM) spectroscopy ( Delta lambda ~1-10 AA). Molybdenum was the dominant high-Z impurity and originated from the molybdenum armour tiles covering all of the plasma facing surfaces (including the inner column, the poloidal divertor plates and the ion cyclotron resonant frequency (ICRF) limiter) at Alcator C-Mod. Despite the all metal first wall, a carbon concentration of 1 to 2% existed in the plasma and was the major low-Z impurity in Alcator C-Mod. Thus, the behaviour of intrinsic molybdenum and carbon penetrating into the main plasma and the effect on the plasma must be measured and characterized during various modes of Alcator C-Mod operation. To this end, soft X-ray extreme ultraviolet (XUV) emission lines of charge states, ranging from hydrogen-like to helium-like lines of carbon (radius/minor radius, r/a~1) at the plasma edge to potassium to chlorine-like (0.4

Numerical investigation of non-perturbative kinetic effects of energetic particles on toroidicity-induced Alfvén eigenmodes in tokamaks and stellarators

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

Slaby, Christoph; Könies, Axel; Kleiber, Ralf

2016-09-15

The resonant interaction of shear Alfvén waves with energetic particles is investigated numerically in tokamak and stellarator geometry using a non-perturbative MHD-kinetic hybrid approach. The focus lies on toroidicity-induced Alfvén eigenmodes (TAEs), which are most easily destabilized by a fast-particle population in fusion plasmas. While the background plasma is treated within the framework of an ideal-MHD theory, the drive of the fast particles, as well as Landau damping of the background plasma, is modelled using the drift-kinetic Vlasov equation without collisions. Building on analytical theory, a fast numerical tool, STAE-K, has been developed to solve the resulting eigenvalue problem usingmore » a Riccati shooting method. The code, which can be used for parameter scans, is applied to tokamaks and the stellarator Wendelstein 7-X. High energetic-ion pressure leads to large growth rates of the TAEs and to their conversion into kinetically modified TAEs and kinetic Alfvén waves via continuum interaction. To better understand the physics of this conversion mechanism, the connections between TAEs and the shear Alfvén wave continuum are examined. It is shown that, when energetic particles are present, the continuum deforms substantially and the TAE frequency can leave the continuum gap. The interaction of the TAE with the continuum leads to singularities in the eigenfunctions. To further advance the physical model and also to eliminate the MHD continuum together with the singularities in the eigenfunctions, a fourth-order term connected to radiative damping has been included. The radiative damping term is connected to non-ideal effects of the bulk plasma and introduces higher-order derivatives to the model. Thus, it has the potential to substantially change the nature of the solution. For the first time, the fast-particle drive, Landau damping, continuum damping, and radiative damping have been modelled together in tokamak- as well as in stellarator geometry.« less

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.

What`s fair is fair

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

Nachtrieb, R.; Freidberg, J.P.

The newly elucidated strategy for the magnetic fusion program set forth by the Department of Energy calls for increased emphasis on alternate concepts. This strategy is motivated by the recognition that in spite of its many attractive features, a tokamak tends to be a low power density device, ultimately translating into large and corresponding expensive reactor. ITER, as it is currently envisaged, is a good example of a large, expensive, plain vanilla tokamak. In its defense, ITER rightly claims that its base design is very conservative in order to minimize the risk of failure. In order to increase power densitymore » and reduce cost there are two qualitatively different approaches that one can follow: discover advanced modes of tokamak operation or develop near alternate concepts. To decide which path to follow is a difficult task because of the uncertainties involved in making accurate comparisons between different concepts at different stages of development. One area, however, that most would agree is meaningful is ideal MHD stability. For any given concept to be credible as a reactor, it must at least be stable against macroscopic ideal MHD modes. The TPX design, for instance, goes to considerable trouble to obtain stability against external kinks: a close fitting metallic cage, rotation to stabilize the resistive wall version of the external kink, and, if all else fails, feedback. For credibility any other advanced tokamak or alternate concept should be held to the same standards of ideal MHD stability. As a first step in addressing this requirement we have investigated the stability of the RFP since it can be simply and accurately modeled as a straight cylinder. The RFP is well known to have good stability at high P against internal modes but is very unstable to external modes. We have developed a linear stability code which treats the plasma as an ideal compressible fluid, and includes longitudinal flow and a resistive wall.« less

Electron Temperature and Density in Local Helicity Injection and High betat Plasmas

NASA Astrophysics Data System (ADS)

Schlossberg, David J.

Tokamak startup in a spherical torus (ST) and an ST-based fusion nuclear science facility can greatly benefit from using non-inductive methods. The Pegasus Toroidal Experiment has developed a non-inductive startup technique using local helicity injection (LHI). Electron temperature, T e(r), and density, ne( r), profiles during LHI are unknown. These profiles are critical for understanding both the physics of the injection and relaxation mechanisms, as well as for extrapolating this technique to larger devices. A new Thomson scattering system has been designed, installed, and used to characterize Te(r, t) and ne(r, t) during LHI. The diagnostic leverages new technology in image intensified CCD cameras, high-efficiency diffraction gratings, and reliable Nd:YAG lasers. Custom systems for stray light mitigation, fast shuttering, and precision timing have been developed and implemented. The overall system provides a low-maintenance, economic, and effective means to explore novel physics regimes in Pegasus. Electron temperature and density profiles during LHI have been measured for the first time. Results indicate Te(r) peaked in the core of plasmas, and sustained while plasmas are coupled to injection drive. Electron densities also peak near the core of the tokamak, up to local values of n e ˜ 1.5 x 1019 m -3. A comparison of Te( r, t) has been made between discharges with dominant drive voltage from induction versus helicity injection. In both cases Te ( r, t) profiles remain peaked, with values for Te ,max > 150 eV in dominantly helicity-driven plasmas using high-field side LHI. Sustained values of betat ˜ 100% have been demonstrated in a tokamak for the first time. Plasmas are created and driven entirely non-solenoidally, and exhibit MHD stability. Measured temperature and density profiles are used to constrain magnetic equilibrium reconstructions, which calculate 80% < betat < 100% throughout a toroidal field ramp-down. For a continued decrease in the toroidal field these plasmas disrupt near the ideal MHD no-wall stability limit predicted by the DCON code. Mode analyses of predicted and measured MHD agree, and suggest discharges terminate by an intermediate-m, n=1 external mode. A localized region of minimum |B| has been identified in these discharges, and modeling shows access to it depends on both plasma pressure and magnetic geometry. This magnetic well is shown to persist over several milliseconds, in both constant toroidal field and ramp-down cases.

Saturation of Alfvén modes in tokamaks

DOE PAGES

White, Roscoe; Gorelenkov, Nikolai; Gorelenkova, Marina; ...

2016-09-20

Here, the growth of Alfvén modes driven unstable by a distribution of high energy particles up to saturation is investigated with a guiding center code, using numerical eigenfunctions produced by linear theory and a numerical high energy particle distribution, in order to make detailed comparison with experiment and with models for saturation amplitudes and the modification of beam profiles. Two innovations are introduced. First, a very noise free means of obtaining the mode-particle energy and momentum transfer is introduced, and secondly, a spline representation of the actual beam particle distribution is used.

Saturation of Alfvén modes in tokamaks

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

White, Roscoe; Gorelenkov, Nikolai; Gorelenkova, Marina

Here, the growth of Alfvén modes driven unstable by a distribution of high energy particles up to saturation is investigated with a guiding center code, using numerical eigenfunctions produced by linear theory and a numerical high energy particle distribution, in order to make detailed comparison with experiment and with models for saturation amplitudes and the modification of beam profiles. Two innovations are introduced. First, a very noise free means of obtaining the mode-particle energy and momentum transfer is introduced, and secondly, a spline representation of the actual beam particle distribution is used.

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.

Numerical optimization of three-dimensional coils for NSTX-U

DOE PAGES

Lazerson, S. A.; Park, J. -K.; Logan, N.; ...

2015-09-03

A tool for the calculation of optimal three-dimensional (3D) perturbative magnetic fields in tokamaks has been developed. The IPECOPT code builds upon the stellarator optimization code STELLOPT to allow for optimization of linear ideal magnetohydrodynamic perturbed equilibrium (IPEC). This tool has been applied to NSTX-U equilibria, addressing which fields are the most effective at driving NTV torques. The NTV torque calculation is performed by the PENT code. Optimization of the normal field spectrum shows that fields with n = 1 character can drive a large core torque. It is also shown that fields with n = 3 features are capablemore » of driving edge torque and some core torque. Coil current optimization (using the planned in-vessel and existing RWM coils) on NSTX-U suggest the planned coils set is adequate for core and edge torque control. In conclusion, comparison between error field correction experiments on DIII-D and the optimizer show good agreement.« less

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.

Photon Throughput Calculations for a Spherical Crystal Spectrometer

NASA Astrophysics Data System (ADS)

Gilman, C. J.; Bitter, M.; Delgado-Aparicio, L.; Efthimion, P. C.; Hill, K.; Kraus, B.; Gao, L.; Pablant, N.

2017-10-01

X-ray imaging crystal spectrometers of the type described in Refs. have become a standard diagnostic for Doppler measurements of profiles of the ion temperature and the plasma flow velocities in magnetically confined, hot fusion plasmas. These instruments have by now been implemented on major tokamak and stellarator experiments in Korea, China, Japan, and Germany and are currently also being designed by PPPL for ITER. A still missing part in the present data analysis is an efficient code for photon throughput calculations to evaluate the chord-integrated spectral data. The existing ray tracing codes cannot be used for a data analysis between shots, since they require extensive and time consuming numerical calculations. Here, we present a detailed analysis of the geometrical properties of the ray pattern. This method allows us to minimize the extent of numerical calculations and to create a more efficient code. This work was performed under the auspices of the U.S. Department of Energy by Princeton Plasma Physics Laboratory under contract DE-AC02-09CH11466.

Design and Manufacturing of the Kstar Tokamak Helium Refrigeration System

NASA Astrophysics Data System (ADS)

Dauguet, P.; Briend, P.; Abe, I.; Fauve, E.; Bernhardt, J. M.; Andrieu, F.; Beauvisage, J.

2008-03-01

The KSTAR (Korean Superconducting Tokamak Advanced Research) project makes intensive use of superconducting (SC) magnets operated at 4.4 K. The cold components of KSTAR require a forced flow of supercritical helium for magnets and structure, boiling liquid helium for current leads, and gaseous helium for thermal shields. A helium refrigeration system has been custom-designed for this project. The purpose of this paper is to give a brief overview of the proposed cryogenic system. The specified thermal loads for the different operating modes are presented. This specification results in the definition of a design mode for the refrigerator. The design and construction of the resulting 9 kW at 4.5-K Helium Refrigeration System (HSR) are presented.

Quasistationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E × B Flow During High Performance DIII-D Tokamak Discharges [A New, Quasi-stationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E × B Flow During High Performance DIII-D Tokamak Discharges

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

Barada, Kshitish; Rhodes, Terry L.; Burrell, Keith H.

A new, long-lived limit cycle oscillation (LCO) regime has been observed in the edge of near zero torque high-performance DIII-D tokamak plasma discharges. These LCOs are localized and comprised of density turbulence, gradient drives, and E X B velocity shear damping ( E and B are the local radial electric and total magnetic fields). Density turbulence sequentially acts as a predator (via turbulence transport) of profile gradients and a prey (via shear suppression) to the E X B velocity shear. Reported here for the first time, a unique spatiotemporal variation of the local E X B velocity which is foundmore » to be essential for the existence of this system. The LCO system is quasi-stationary, existing from 3 to 12 plasma energy confinement times (~30 to 900 LCO cycles) limited by hardware constraints. In conclusion, this plasma system appears to contribute strongly to the edge transport in these high-performance and transient-free plasmas as evident from oscillations in transport relevant edge parameters at LCO timescale.« less

A high speed compact microwave interferometer for density fluctuation measurements in Sino-UNIted Spherical Tokamak

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

Zhong, H., E-mail: zhongh14@126.com; Tan, Y.; Liu, Y. Q.

2016-11-15

A single-channel 3 mm interferometer has been developed for plasma density diagnostics in the Sino-UNIted Spherical Tokamak (SUNIST). The extremely compact microwave interferometer utilizes one corrugated feed horn antenna for both emitting and receiving the microwave. The beam path lies on the equatorial plane so the system would not suffer from beam path deflection problems due to the symmetry of the cross section. A focusing lens group and an oblique vacuum window are carefully designed to boost the signal to noise ratio, which allows this system to show good performance even with the small-diameter central column itself as a reflector,more » without a concave mirror. The whole system discards the reference leg for maximum compactness, which is particularly suitable for the small-sized tokamak. An auto-correcting algorithm is developed to calculate the phase evolution, and the result displays good phase stability of the whole system. The intermediate frequency is adjustable and can reach its full potential of 2 MHz for best temporal resolution. Multiple measurements during ohmic discharges proved the interferometer’s capability to track typical density fluctuations in SUNIST, which enables this system to be utilized in the study of MHD activities.« less

Quasistationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E × B Flow During High Performance DIII-D Tokamak Discharges [A New, Quasi-stationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E × B Flow During High Performance DIII-D Tokamak Discharges

DOE PAGES

Barada, Kshitish; Rhodes, Terry L.; Burrell, Keith H.; ...

2018-03-27

A new, long-lived limit cycle oscillation (LCO) regime has been observed in the edge of near zero torque high-performance DIII-D tokamak plasma discharges. These LCOs are localized and comprised of density turbulence, gradient drives, and E X B velocity shear damping ( E and B are the local radial electric and total magnetic fields). Density turbulence sequentially acts as a predator (via turbulence transport) of profile gradients and a prey (via shear suppression) to the E X B velocity shear. Reported here for the first time, a unique spatiotemporal variation of the local E X B velocity which is foundmore » to be essential for the existence of this system. The LCO system is quasi-stationary, existing from 3 to 12 plasma energy confinement times (~30 to 900 LCO cycles) limited by hardware constraints. In conclusion, this plasma system appears to contribute strongly to the edge transport in these high-performance and transient-free plasmas as evident from oscillations in transport relevant edge parameters at LCO timescale.« less

Conceptual design and proof-of-principle testing of the real-time multispectral imaging system MANTIS

NASA Astrophysics Data System (ADS)

Vijvers, W. A. J.; Mumgaard, R. T.; Andrebe, Y.; Classen, I. G. J.; Duval, B. P.; Lipschultz, B.

2017-12-01

The Multispectral Advanced Narrowband Tokamak Imaging System (MANTIS) is proposed to resolve the steep temperature and density gradients in the scrape-off layer of tokamaks in real-time. The initial design is to deliver two-dimensional distributions of key plasma parameters of the TCV tokamak to a real-time control system in order to enable novel control strategies, while providing new insights into power exhaust physics in the full offline analysis. This paper presents the conceptual system design, the mechanical and optical design of a prototype that was built to assess the optical performance, and the results of the first proof-of-principle tests of the prototype. These demonstrate a central resolving power of 50-46 line pairs per millimeter (CTF50) in the first four channels. For the additional channels, the sharpness is a factor two worse for the odd channels (likely affected by sub-optimal alignment), while the even channels continue the trend observed for the first four channels of 3% degradation per channel. This is explained by the self-cancellation of off-axis aberrations, which is an attractive property of the chosen optical design. The results show that at least a 10-channel real-time multispectral imaging system is feasible.

Stress and Thermal Analysis of the In-Vessel Resonant Magnetic Perturbation Coils on the J-TEXT Tokamak

NASA Astrophysics Data System (ADS)

Hao, Changduan; Zhang, Ming; Ding, Yonghua; Rao, Bo; Cen, Yishun; Zhuang, Ge

2012-01-01

A set of four in-vessel saddle coils was designed to generate a helical field on the J-TEXT tokamak to study the influences of the external perturbation field on plasma. The coils are fed with alternating current up to 10 kA at frequency up to 10 kHz. Due to the special structure, complex thermal environment and limited space in the vacuum chamber, it is very important to make sure that the coils will not be damaged when undergoing the huge electromagnetic forces in the strong toroidal field, and that their temperatures don't rise too much and destroy the insulation. A 3D finite element model is developed in this paper using the ANSYS code, stresses are analyzed to find the worst condition, and a mounting method is then established. The results of the stress and modal analyses show that the mounting method meets the strength requirements. Finally, a thermal analysis is performed to study the cooling process and the temperature distribution of the coils.

Disruption Event Characterization and Forecasting in Tokamaks

NASA Astrophysics Data System (ADS)

Berkery, J. W.; Sabbagh, S. A.; Park, Y. S.; Ahn, J. H.; Jiang, Y.; Riquezes, J. D.; Gerhardt, S. P.; Myers, C. E.

2017-10-01

The Disruption Event Characterization and Forecasting (DECAF) code, being developed to meet the challenging goal of high reliability disruption prediction in tokamaks, automates data analysis to determine chains of events that lead to disruptions and to forecast their evolution. The relative timing of magnetohydrodynamic modes and other events including plasma vertical displacement, loss of boundary control, proximity to density limits, reduction of safety factor, and mismatch of the measured and desired plasma current are considered. NSTX/-U databases are examined with analysis expanding to DIII-D, KSTAR, and TCV. Characterization of tearing modes has determined mode bifurcation frequency and locking points. In an NSTX database exhibiting unstable resistive wall modes (RWM), the RWM event and loss of boundary control event were found in 100%, and the vertical displacement event in over 90% of cases. A reduced kinetic RWM stability physics model is evaluated to determine the proximity of discharges to marginal stability. The model shows high success as a disruption predictor (greater than 85%) with relatively low false positive rate. Supported by US DOE Contracts DE-FG02-99ER54524, DE-AC02-09CH11466, and DE-SC0016614.

Toroidal Ampere-Faraday Equations Solved Simultaneously with CQL3D Fokker-Planck Time-Evolution

NASA Astrophysics Data System (ADS)

Harvey, R. W. (Bob); Petrov, Yu. V. (Yuri); Forest, C. B.; La Haye, R. J.

2017-10-01

A self-consistent, time-dependent toroidal electric field calculation is a key feature of a complete 3D Fokker-Planck kinetic distribution radial transport code for f(v,theta,rho,t). We discuss benchmarking and first applications of an implementation of the Ampere-Faraday equation for the self-consistent toroidal electric field, as applied to (1) resistive turn on of applied electron cyclotron current in the DIII-D tokamak giving initial back current adjacent to the direct CD region and having possible NTM stabilization implications, and (2) runaway electron production in tokamaks due to rapid reduction of the plasma temperature as occurs in pellet injection, massive gas injection, or a plasma disruption. Our previous results assuming a constant current density (Lenz' Law) model showed that prompt ``hot-tail runaways'' dominated ``knock-on'' and Dreicer ``drizzle'' runaways; we perform full-radius modeling and examine modifications due to the more complete Ampere-Faraday solution. Presently, the implementation relies on a fixed shape eqdsk, and this limitation will be addressed in future work. Research supported by USDOE FES award ER54744.

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

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.

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.

Numerical investigation of disruption characteristics for the snowflake divertor configuration in HL-2M

NASA Astrophysics Data System (ADS)

Xue, L.; Duan, X. R.; Zheng, G. Y.; Liu, Y. Q.; Pan, Y. D.; Yan, S. L.; Dokuka, V. N.; Lukash, V. E.; Khayrutdinov, R. R.

2016-05-01

Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench.

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.

Toroidal ripple transport of beam ions in the mega-ampere spherical tokamak

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

McClements, K. G.; Hole, M. J.

The transport of injected beam ions due to toroidal magnetic field ripple in the mega-ampere spherical tokamak (MAST) is quantified using a full orbit particle tracking code, with collisional slowing-down and pitch-angle scattering by electrons and bulk ions taken into account. It is shown that the level of ripple losses is generally rather low, although it depends sensitively on the major radius of the outer midplane plasma edge; for typical values of this parameter in MAST plasmas, the reduction in beam heating power due specifically to ripple transport is less than 1%, and the ripple contribution to beam ion diffusivitymore » is of the order of 0.1 m{sup 2} s{sup -1} or less. It is concluded that ripple effects make only a small contribution to anomalous transport rates that have been invoked to account for measured neutron rates and plasma stored energies in some MAST discharges. Delayed (non-prompt) losses are shown to occur close to the outer midplane, suggesting that banana-drift diffusion is the most likely cause of the ripple-induced losses.« less

Calculation of ion distribution functions and neoclassical transport in the edge of single-null divertor tokamaks

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Cohen, R. H.; Xu, X. Q.

2007-11-01

The ion distribution function in the H-mode pedestal region and outward across the magnetic separatrix is expected to have a substantial non-Maxwellian character owing to the large banana orbits and steep gradients in temperature and density. The 4D (2r,2v) version of the TEMPEST continuum gyrokinetic code is used with a Coulomb collision model to calculate the ion distribution in a single-null tokamak geometry throughout the pedestal/scrape-off-layer regions. The mean density, parallel velocity, and energy radial profiles are shown at various poloidal locations. The collisions cause neoclassical energy transport through the pedestal that is then lost to the divertor plates along the open field lines outside the separatrix. The resulting heat flux profiles at the inner and outer divertor plates are presented and discussed, including asymmetries that depend on the B-field direction. Of particular focus is the effect on ion profiles and fluxes of a radial electric field exhibiting a deep well just inside the separatrix, which reduces the width of the banana orbits by the well-known squeezing effect.

[ital n]=5 to [ital n]=5 soft-x-ray emission of uranium in a high-temperature low-density tokamak plasma

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

Fournier, K.B.; Finkenthal, M.; Lippmann, S.

1994-11-01

The soft-x-ray uranium emission in the 60--200-A range recorded from a high-temperature ([similar to]1 keV) low-density ([similar to]10[sup 13] cm[sup [minus]3]) tokamak plasma has been analyzed by comparison with theoretical level structure and line-intensity calculations. In an extension of previous work [Finkenthal [ital et] [ital al]., Phys. Rev. A 45, 5846 (1992)], theoretical U XXV, U XXX, U XXXI, and U XXXII [ital n]=5 to [ital n]=5 spectra have been computed for the relevant plasma parameters. Fully relativistic parametric potential computer codes have been used for the [ital ab] [ital initio] atomic-structure calculations, and electron-impact excitation rates have been computedmore » in the distorted-wave approximation. 5[ital s]-5[ital p] spectral lines and quasicontinua of U XXX, U XXXI, and U XXXII are identified in the 165--200-A wavelength band. An unambiguous line identification is hampered by theoretical uncertainties and the blending of emission from adjacent charge states.« less

Steady State Advanced Tokamak (SSAT): The mission and the machine

NASA Astrophysics Data System (ADS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the U.S. National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new 'Steady State Advanced Tokamak' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO.

Gas Fuelling System for SST-1Tokamak

NASA Astrophysics Data System (ADS)

Dhanani, Kalpesh; Raval, D. C.; Khan, Ziauddin; Semwal, Pratibha; George, Siju; Paravastu, Yuvakiran; Thankey, Prashant; Khan, M. S.; Pradhan, Subrata

2017-04-01

SST-1 Tokamak, the first Indian Steady-state Superconducting experimental device is at present under operation in the Institute for Plasma Research. For plasma break down & initiation, piezoelectric valve based gas feed system is implemented as a primary requirement due to its precise control, easy handling, low construction and maintenance cost and its flexibility in the selection of the working gas. Hydrogen gas feeding with piezoelectric valve is used in the SST-1 plasma experiments. The piezoelectric valves used in SST-1 are remotely driven by a PXI based platform and are calibrated before each SST-1 plasma operation with precise control. This paper will present the technical development and the results of the gas fuelling system of SST-1.

Stability at high performance in the MAST spherical tokamak

NASA Astrophysics Data System (ADS)

Buttery, R. J.; Akers, R.; Arends, E.; Conway, N. J.; Counsell, G. F.; Cunningham, G.; Gimblett, C. G.; Gryaznevich, M.; Hastie, R. J.; Hole, M. J.; Lehane, I.; Martin, R.; Patel, A.; Pinfold, T.; Sauter, O.; Taylor, D.; Turri, G.; Valovic, M.; Walsh, M. J.; Wilson, H. R.; MAST Team

2004-09-01

The development of reliable H-modes on MAST, together with advances in heating power and a range of high spatial resolution diagnostics, has provided a platform to enable MAST to address some of the most important issues of tokamak stability. In particular the high bgr potential of the spherical tokamak is highlighted with stable operation at bgrN ~ 5-6, bgrT ~ 16% and bgrp up to ~2. Magnetic diagnostic evaluation of the global bgr parameters is independently confirmed by kinetic profile data. Calculations indicate that the bgrN values are in the vicinity of no-wall stability limits. Studies of neoclassical tearing modes (NTMs) have been extended to explore their effects and develop avoidance strategies. Experiments have demonstrated that sawteeth play a strong role in triggering NTMs—by avoiding large sawteeth a much higher bgrN value has been reached. The significance of NTMs is confirmed, with large islands observed using the 300 point Thomson scattering diagnostic, and locking of large n = 1 modes frequently leading to disruptions, which become more rapid at low q95. The role of error fields has been explored. H-mode plasmas are also limited by edge localized modes (ELMs), with confinement degraded as the ELM frequency rises. However, in contrast to the conventional tokamak, the ELMs in high performing regimes on MAST (HIPB98Y2 ~ 1) appear to be type III in nature. Modelling using the ELITE code, which incorporates finite n corrections, identifies instability to peeling modes, consistent with a type III interpretation. It also shows considerable scope to raise pressure gradients before ballooning type modes (perhaps associated with type I ELMs) occur. The calculations show that narrow pedestals can support much stronger pressure gradients than might be expected from simple n = infin ballooning calculations. Finally sawteeth are shown to degrade confinement by ~10-15% in particular cases examined. They are observed not to remove the q = 1 surface in the cases where snakes are present—various physics models of the sawteeth are now being explored. Thus research on MAST is not only demonstrating stable operation at high performance levels and developing methods to control instabilities; it is also providing detailed tests of the stability physics and models applicable to conventional tokamaks, such as ITER.

A lithium deposition system for tokamak devices*

NASA Astrophysics Data System (ADS)

Graziul, Christopher; Majeski, Richard; Kaita, Robert; Hoffman, Daniel; Timberlake, John; Card, David

2002-11-01

The production of a lithium deposition system using commercially available components is discussed. This system is intended to provide a fresh lithium wall coating between discharges in a tokamak. For this purpose, a film 100-200 Å thick is sufficient to ensure that the plasma interacts solely with the lithium. A test system consisting of a lithium evaporator and a deposition monitor has been designed and constructed to investigate deposition rates and coverage. A Thermionics 3kW e-gun is used to rapidly evaporate small amounts of solid lithium. An Inficon XTM/2 quartz deposition monitor then measures deposition rate at varying distances, positions and angles relative to the e-gun crucible. Initial results from the test system will be presented. *Supported by US DOE contract #DE-AC02-76CH-03073

Data acquisition and processing system for the HT-6M tokamak fusion experiment

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

Shu, Y.T.; Liu, G.C.; Pang, J.Q.

1987-08-01

This paper describes a high-speed data acquisition and processing system which has been successfully operated on the HT-6M tokamak fusion experimental device. The system collects, archives and analyzes up to 512 kilobytes of data from each shot of the experiment. A shot lasts 50-150 milliseconds and occurs every 5-10 minutes. The system consists of two PDP-11/24 computer systems. One PDP-11/24 is used for real-time data taking and on-line data analysis. It is based upon five CAMAC crates organized into a parallel branch. Another PDP-11/24 is used for off-line data processing. Both data acquisition software RSX-DAS and data processing software RSX-DAPmore » have modular, multi-tasking and concurrent processing features.« less

Status of the tokamak program

NASA Astrophysics Data System (ADS)

Sheffield, J.

1981-08-01

For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

Improving spatial and spectral resolution of TCV Thomson scattering

NASA Astrophysics Data System (ADS)

Hawke, J.; Andrebe, Y.; Bertizzolo, R.; Blanchard, P.; Chavan, R.; Decker, J.; Duval, B.; Lavanchy, P.; Llobet, X.; Marlétaz, B.; Marmillod, P.; Pochon, G.; Toussaint, M.

2017-12-01

The recently completed MST2 upgrade to the Thomson scattering (TS) system on TCV (Tokamak à Configuration Variable) at the Swiss Plasma Center aims to provide an enhanced spatial and spectral resolution while maintaining the high level of diagnostic flexibility for the study of TCV plasmas. The MST2 (Medium Sized Tokamak) is a work program within the Eurofusion ITER physics department, aimed at exploiting Europe's medium sized tokamak programs for a better understanding of ITER physics. This upgrade to the TCV Thomson scattering system involved the installation of 40 new compact 5-channel spectrometers and modifications to the diagnostics fiber optic design. The complete redesign of the fiber optic backplane incorporates fewer larger diameter fibers, allowing for a higher resolution in both the core and edge of TCV plasmas along the laser line, with a slight decrease in the signal to noise ratio of Thomson measurements. The 40 new spectrometers added to the system are designed to cover the full range of temperatures expected in TCV, able to measure electron temperatures (Te) with high precision between (6 eV and 20 keV) . The design of these compact spectrometers stems originally from the design utilized in the MAST (Mega Amp Spherical Tokamak) TS system located in Oxfordshire, United Kingdom. This design was implemented on TCV with an overall layout of optical fibers and spectrometers to achieve an overall increase in the spatial resolution, specifically a resolution of approximately 1% of the minor radius within the plasma pedestal region. These spectrometers also enhance the diagnostic spectral resolution, especially within the plasma edge, due to the low Te measurement capabilities. These additional spectrometers allow for a much greater diagnostic flexibility, allowing for quality full Thomson profiles in 75% of TCV plasma configurations.

High resolution main-ion charge exchange spectroscopy in the DIII-D H-mode pedestal.

PubMed

Grierson, B A; Burrell, K H; Chrystal, C; Groebner, R J; Haskey, S R; Kaplan, D H

2016-11-01

A new high spatial resolution main-ion (deuterium) charge-exchange spectroscopy system covering the tokamak boundary region has been installed on the DIII-D tokamak. Sixteen new edge main-ion charge-exchange recombination sightlines have been combined with nineteen impurity sightlines in a tangentially viewing geometry on the DIII-D midplane with an interleaving design that achieves 8 mm inter-channel radial resolution for detailed profiles of main-ion temperature, velocity, charge-exchange emission, and neutral beam emission. At the plasma boundary, we find a strong enhancement of the main-ion toroidal velocity that exceeds the impurity velocity by a factor of two. The unique combination of experimentally measured main-ion and impurity profiles provides a powerful quasi-neutrality constraint for reconstruction of tokamak H-mode pedestals.

The Spherical Tokamak MEDUSA for Mexico

NASA Astrophysics Data System (ADS)

Ribeiro, C.; Salvador, M.; Gonzalez, J.; Munoz, O.; Tapia, A.; Arredondo, V.; Chavez, R.; Nieto, A.; Gonzalez, J.; Garza, A.; Estrada, I.; Jasso, E.; Acosta, C.; Briones, C.; Cavazos, G.; Martinez, J.; Morones, J.; Almaguer, J.; Fonck, R.

2011-10-01

The former spherical tokamak MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14m, a < 0.10m, BT < 0.5T, Ip < 40kA, 3ms pulse) is currently being recomissioned at the Universidad Autónoma de Nuevo León, Mexico, as part of an agreement between the Faculties of Mech.-Elect. Eng. and Phy. Sci.-Maths. The main objective for having MEDUSA is to train students in plasma physics & technical related issues, aiming a full design of a medium size device (e.g. Tokamak-T). Details of technical modifications and a preliminary scientific programme will be presented. MEDUSA-MX will also benefit any developments in the existing Mexican Fusion Network. Strong liaison within national and international plasma physics communities is expected. New activities on plasma & engineering modeling are expected to be developed in parallel by using the existing facilities such as a multi-platform computer (Silicon Graphics Altix XE250, 128G RAM, 3.7TB HD, 2.7GHz, quad-core processor), ancillary graph system (NVIDIA Quadro FE 2000/1GB GDDR-5 PCI X16 128, 3.2GHz), and COMSOL Multiphysics-Solid Works programs.

The Development of High-Intensity Negative Ion Sources and Beams in the USSR

DTIC Science & Technology

1981-09-01

ion beams as the basis for creating neutral beams for injection into mirror traps and tokamaks, for inertial confinement fusion, and possibly for...create intense neutral beams for injection systems for mirror traps and tokamaks and for inertial confinement fusion. These applications require high...Scient. Instr., Vol. 44, 1973, p. 145. 46. Gabovich, M. D., Yu. N. Kozyrev , A. P. Nayda, L. S. Simonenko, I. A. Soloshenko, "H- Ion Beam Limit from a

SOVRaD - A Digest of Recent Soviet R and D Articles. Volume 2, Number 6, 1976

DTIC Science & Technology

1976-06-01

6 Laser- Powered Rocket Model 1 High- Power CO2 Laser Radiation Effect in SF6 1 Tests With 9-Beam Laser Fusion Systems 1 Focusing Optics For...Boundary Layer 6 Deformation Theory of Artif.cial Muscles . 6 Dolphin Swimming Stereophotogrammetry 7 Stable Spark Gap for High- Power Pulsers 7...8 Resume of Soviet Tokamak Program .............. 9 First Measurements of Tokamak-10 Plasma , . . 10 Electrochemical Power Generation 11

Spherical tokamaks with plasma centre-post

NASA Astrophysics Data System (ADS)

Ribeiro, Celso

2013-10-01

The metal centre-post (MCP) in tokamaks is a structure which carries the total toroidal field current and also houses the Ohmic heating solenoid in conventional or low aspect ratio (Spherical)(ST) tokamaks. The MCP and solenoid are critical components for producing the toroidal field and for the limited Ohmic flux in STs. Constraints for a ST reactor related to these limitations lead to a minimum plasma aspect ratio of 1.4 which reduces the benefit of operation at higher betas in a more compact ST reactor. Replacing the MCP is of great interest for reactor-based ST studies since the device is simplified, compactness increased, and maintenance reduced. An experiment to show the feasibility of using a plasma centre-post (PCP) is being currently under construction and involves a high level of complexity. A preliminary study of a very simple PCP, which is ECR(Electron Cyclotron Resonance)-assisted and which includes an innovative fuelling system based on pellet injection, has recently been reported. This is highly suitable for an ultra-low aspect ratio tokamak (ULART) device. Advances on this PCP ECR-assisted concept within a ULART and the associated fuelling system are presented here, and will include the field topology for the PCP ECR-assisted scheme, pellet ablation modeling, and a possible global equilibrium simulation. VIE-ITCR, IAEA-CRP contr.17592, National Instruments-Costa Rica.

Current drive by spheromak injection into a tokamak

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

Brown, M.R.; Bellan, P.M.

1990-04-30

We report the first observation of current drive by injection of a spheromak plasma into a tokamak (Caltech ENCORE small reasearch tokamak) due to the process of helicity injection. After an abrupt 30% increase, the tokamak current decays by a factor of 3 due to plasma cooling caused by the merging of the relatively cold spheromak with the tokamak. The tokamak density profile peaks sharply due to the injected spheromak plasma ({ital {bar n}}{sub 3} increases by a factor of 6) then becomes hollow, suggestive of an interchange instability.

Design and application of a new control system for tokamak ECRH power supply

NASA Astrophysics Data System (ADS)

Hao, Xu; Zhang, Jian; Huang, Yiyun

2016-03-01

The biggest challenge of designing and building tokamak electron cyclotron resonance heating (ECRH) pulse step modulation (PSM) power supply is satisfying its required output voltage rising time to be less than 100 µs while suppressing the voltage overshoot to be no more than 1%. To fulfill the two requirements, a new control strategy with startup time in microsecond range is proposed in this paper, and a new control system to realize the control strategy is introduced. The control system was built and tested on 60 kV/50 A ECRH power supply. The experimental results indicate that the control system can restrain the overshoot effectively, increase response speed, and obviously improve the dynamic characteristics of the PSM power supply system. Thus, the proposed control system helps the PSM power supply to meet the design specifications.

A Distributed Synchronization and Timing System on the EAST Tokamak

NASA Astrophysics Data System (ADS)

Luo, Jiarong; Wu, Yichun; Shu, Yantai

2008-08-01

A key requirement for the EAST distributed control system (EASTDCS) is time synchronization to an accuracy of <1 mus. In 2006 a Distributed Synchronization and Timing System (DSTS) was set up, which is based on the ATmega128 AVR microcontroller and the Nut/OS embedded Real Time Operating System (RTOS). The DSTS provides the control and the data acquisition systems with reference clocks (0.01 Hz 10 MHz) and delayed trigger times ( 1 mus 4294 s). These are produced by a Core Module Unit (CMU) connected by optical fibres to many Local Synchronized Node Units (LSNU). The fibres provide immunity from electrical noise and are of equal length to match clock and trigger delays between systems. This paper describes the architecture of the DSTS on the EAST tokamak and provides an overview of the characteristics of the main and local units.

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.

Spectroscopic investigation of carbon migration with tungsten walls in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Dux, R.; Harhausen, J.; Maggi, C. F.; Neu, R.; Pütterich, T.; Rohde, V.; Schmid, K.; Wolfrum, E.; ASDEX Upgrade Team

2007-06-01

Spectroscopic measurements of carbon fluxes in the mainly tungsten-coated ASDEX Upgrade tokamak are analysed with a particle transport and migration code. The transport parameters for deuterium and carbon are calibrated against flux measurements for different experimental conditions. Additional information is obtained from the re-appearance time of carbon after a boronisation. The code reproduces the experimental finding that despite a 85% (2006 campaign) tungsten coverage of the primary PFCs, the carbon concentration in the core and edge plasma is reduced by about a factor 2 only compared to full carbon PFCs. This behaviour is explained by the strong main chamber recycling of carbon in comparison with the loss flux to the inner divertor. The quick recovery of the carbon level in the plasma after a boronisation is explained by carbon influx from the outer divertor.

Automation of the guiding center expansion

NASA Astrophysics Data System (ADS)

Burby, J. W.; Squire, J.; Qin, H.

2013-07-01

We report on the use of the recently developed Mathematica package VEST (Vector Einstein Summation Tools) to automatically derive the guiding center transformation. Our Mathematica code employs a recursive procedure to derive the transformation order-by-order. This procedure has several novel features. (1) It is designed to allow the user to easily explore the guiding center transformation's numerous non-unique forms or representations. (2) The procedure proceeds entirely in cartesian position and velocity coordinates, thereby producing manifestly gyrogauge invariant results; the commonly used perpendicular unit vector fields e1,e2 are never even introduced. (3) It is easy to apply in the derivation of higher-order contributions to the guiding center transformation without fear of human error. Our code therefore stands as a useful tool for exploring subtle issues related to the physics of toroidal momentum conservation in tokamaks.

Automation of The Guiding Center Expansion

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

J. W. Burby, J. Squire and H. Qin

2013-03-19

We report on the use of the recently-developed Mathematica package VEST (Vector Einstein Summation Tools) to automatically derive the guiding center transformation. Our Mathematica code employs a recursive procedure to derive the transformation order-by-order. This procedure has several novel features. (1) It is designed to allow the user to easily explore the guiding center transformation's numerous nonunique forms or representations. (2) The procedure proceeds entirely in cartesian position and velocity coordinates, thereby producing manifestly gyrogauge invariant results; the commonly-used perpendicular unit vector fields e1, e2 are never even introduced. (3) It is easy to apply in the derivation of higher-ordermore » contributions to the guiding center transformation without fear of human error. Our code therefore stands as a useful tool for exploring subtle issues related to the physics of toroidal momentum conservation in tokamaks« 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

Silicon drift detector based X-ray spectroscopy diagnostic system for the study of non-thermal electrons at Aditya tokamak.

PubMed

Purohit, S; Joisa, Y S; Raval, J V; Ghosh, J; Tanna, R; Shukla, B K; Bhatt, S B

2014-11-01

Silicon drift detector based X-ray spectrometer diagnostic was developed to study the non-thermal electron for Aditya tokamak plasma. The diagnostic was mounted on a radial mid plane port at the Aditya. The objective of diagnostic includes the estimation of the non-thermal electron temperature for the ohmically heated plasma. Bi-Maxwellian plasma model was adopted for the temperature estimation. Along with that the study of high Z impurity line radiation from the ECR pre-ionization experiments was also aimed. The performance and first experimental results from the new X-ray spectrometer system are presented.

New MHD feedback control schemes using the MARTe framework in RFX-mod

NASA Astrophysics Data System (ADS)

Piron, Chiara; Manduchi, Gabriele; Marrelli, Lionello; Piovesan, Paolo; Zanca, Paolo

2013-10-01

Real-time feedback control of MHD instabilities is a topic of major interest in magnetic thermonuclear fusion, since it allows to optimize a device performance even beyond its stability bounds. The stability properties of different magnetic configurations are important test benches for real-time control systems. RFX-mod, a Reversed Field Pinch experiment that can also operate as a tokamak, is a well suited device to investigate this topic. It is equipped with a sophisticated magnetic feedback system that controls MHD instabilities and error fields by means of 192 active coils and a corresponding grid of sensors. In addition, the RFX-mod control system has recently gained new potentialities thanks to the introduction of the MARTe framework and of a new CPU architecture. These capabilities allow to study new feedback algorithms relevant to both RFP and tokamak operation and to contribute to the debate on the optimal feedback strategy. This work focuses on the design of new feedback schemes. For this purpose new magnetic sensors have been explored, together with new algorithms that refine the de-aliasing computation of the radial sideband harmonics. The comparison of different sensor and feedback strategy performance is described in both RFP and tokamak experiments.

Design of set-point weighting PI{sup λ} + D{sup μ} controller for vertical magnetic flux controller in Damavand tokamak

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

Rasouli, H.; Fatehi, A.

2014-12-15

In this paper, a simple method is presented for tuning weighted PI{sup λ} + D{sup μ} controller parameters based on the pole placement controller of pseudo-second-order fractional systems. One of the advantages of this controller is capability of reducing the disturbance effects and improving response to input, simultaneously. In the following sections, the performance of this controller is evaluated experimentally to control the vertical magnetic flux in Damavand tokamak. For this work, at first a fractional order model is identified using output-error technique in time domain. For various practical experiments, having desired time responses for magnetic flux in Damavand tokamak,more » is vital. To approach this, at first the desired closed loop reference models are obtained based on generalized characteristic ratio assignment method in fractional order systems. After that, for the identified model, a set-point weighting PI{sup λ} + D{sup μ} controller is designed and simulated. Finally, this controller is implemented on digital signal processor control system of the plant to fast/slow control of magnetic flux. The practical results show appropriate performance of this controller.« less

Calculation of continuum damping of Alfvén eigenmodes in tokamak and stellarator equilibria

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

Bowden, G. W.; Hole, M. J.; Könies, A.

2015-09-15

In an ideal magnetohydrodynamic (MHD) plasma, shear Alfvén eigenmodes may experience dissipationless damping due to resonant interaction with the shear Alfvén continuum. This continuum damping can make a significant contribution to the overall growth/decay rate of shear Alfvén eigenmodes, with consequent implications for fast ion transport. One method for calculating continuum damping is to solve the MHD eigenvalue problem over a suitable contour in the complex plane, thereby satisfying the causality condition. Such an approach can be implemented in three-dimensional ideal MHD codes which use the Galerkin method. Analytic functions can be fitted to numerical data for equilibrium quantities inmore » order to determine the value of these quantities along the complex contour. This approach requires less resolution than the established technique of calculating damping as resistivity vanishes and is thus more computationally efficient. The complex contour method has been applied to the three-dimensional finite element ideal MHD Code for Kinetic Alfvén waves. In this paper, we discuss the application of the complex contour technique to calculate the continuum damping of global modes in tokamak as well as torsatron, W7-X and H-1NF stellarator cases. To the authors' knowledge, these stellarator calculations represent the first calculation of continuum damping for eigenmodes in fully three-dimensional equilibria. The continuum damping of global modes in W7-X and H-1NF stellarator configurations investigated is found to depend sensitively on coupling to numerous poloidal and toroidal harmonics.« less

Finite Beta Boundary Magnetic Fields of NCSX

NASA Astrophysics Data System (ADS)

Grossman, A.; Kaiser, T.; Mioduszewski, P.

2004-11-01

The magnetic field between the plasma surface and wall of the National Compact Stellarator (NCSX), which uses quasi-symmetry to combine the best features of the tokamak and stellarator in a configuration of low aspect ratio is mapped via field line tracing in a range of finite beta in which part of the rotational transform is generated by the bootstrap current. We adopt the methodology developed for W7-X, in which an equilibrium solution is computed by an inverse equilibrium solver based on an energy minimizing variational moments code, VMEC2000[1], which solves directly for the shape of the flux surfaces given the external coils and their currents as well as a bootstrap current provided by a separate transport calculation. The VMEC solution and the Biot-Savart vacuum fields are coupled to the magnetic field solver for finite-beta equilibrium (MFBE2001)[2] code to determine the magnetic field on a 3D grid over a computational domain. It is found that the edge plasma is more stellarator-like, with a complex 3D structure, and less like the ordered 2D symmetric structure of a tokamak. The field lines make a transition from ergodically covering a surface to ergodically covering a volume, as the distance from the last closed magnetic surface is increased. The results are compared with the PIES[3] calculations. [1] S.P. Hirshman et al. Comput. Phys. Commun. 43 (1986) 143. [2] E. Strumberger, et al. Nucl. Fusion 42 (2002) 827. [3] A.H. Reiman and H.S. Greenside, Comput. Phys. Commun. 43, 157 (1986).

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

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.

A frequency tunable, eight-channel correlation ECE system for electron temperature turbulence measurements on the DIII-D tokamak

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

Sung, C., E-mail: csung@physics.ucla.edu; Peebles, W. A.; Wannberg, C.

2016-11-15

A new eight-channel correlation electron cyclotron emission diagnostic has recently been installed on the DIII-D tokamak to study both turbulent and coherent electron temperature fluctuations under various plasma conditions and locations. This unique system is designed to cover a broad range of operation space on DIII-D (1.6-2.1 T, detection frequency: 72-108 GHz) via four remotely selected local oscillators (80, 88, 96, and 104 GHz). Eight radial locations are measured simultaneously in a single discharge covering as much as half the minor radius. In this paper, we present design details of the quasi-optical system, the receiver, as well as representative datamore » illustrating operation of the system.« less

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.

Constructing the spectral web of rotating plasmas

NASA Astrophysics Data System (ADS)

Goedbloed, Hans

2012-10-01

Rotating plasmas are ubiquitous in nature. The theory of MHD stability of such plasmas, initiated a long time ago, has severely suffered from the wide spread misunderstanding that it necessarily involves non-self-adjoint operators. It has been shown (J.P. Goedbloed, PPCF 16, 074001, 2011; Goedbloed, Keppens and Poedts, Advanced Magnetohydrodynamics, Cambridge, 2010) that, on the contrary, spectral theory of moving plasmas can be constructed entirely on the basis of energy conservation and self-adjointness of the occurring operators. The spectral web is a further development along this line. It involves the construction of a network of curves in the complex omega-plane associated with the complex complementary energy, which is the energy needed to maintain harmonic time dependence in an open system. Vanishing of that energy, at the intersections of the mentioned curves, yields the eigenvalues of the closed system. This permits to consider the enormous diversity of MHD instabilities of rotating tokamaks, accretion disks about compact objects, and jets emitted from those objects, from a single view point. This will be illustrated with results obtained with a new spectral code (ROC).

Experimental determination of the correlation properties of plasma turbulence using 2D BES systems

NASA Astrophysics Data System (ADS)

Fox, M. F. J.; Field, A. R.; van Wyk, F.; Ghim, Y.-c.; Schekochihin, A. A.; the MAST Team

2017-04-01

A procedure is presented to map from the spatial correlation parameters of a turbulent density field (the radial and binormal correlation lengths and wavenumbers, and the fluctuation amplitude) to correlation parameters that would be measured by a beam emission spectroscopy (BES) diagnostic. The inverse mapping is also derived, which results in resolution criteria for recovering correct correlation parameters, depending on the spatial response of the instrument quantified in terms of point-spread functions (PSFs). Thus, a procedure is presented that allows for a systematic comparison between theoretical predictions and experimental observations. This procedure is illustrated using the Mega-Ampere Spherical Tokamak BES system and the validity of the underlying assumptions is tested on fluctuating density fields generated by direct numerical simulations using the gyrokinetic code GS2. The measurement of the correlation time, by means of the cross-correlation time-delay method, is also investigated and is shown to be sensitive to the fluctuating radial component of velocity, as well as to small variations in the spatial properties of the PSFs.

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.

Design of the high-resolution soft X-ray imaging system on the Joint Texas Experimental Tokamak

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

Li, Jianchao; Ding, Yonghua, E-mail: yhding@mail.hust.edu.cn; Zhang, Xiaoqing

2014-11-15

A new soft X-ray diagnostic system has been designed on the Joint Texas Experimental Tokamak (J-TEXT) aiming to observe and survey the magnetohydrodynamic (MHD) activities. The system consists of five cameras located at the same toroidal position. Each camera has 16 photodiode elements. Three imaging cameras view the internal plasma region (r/a < 0.7) with a spatial resolution about 2 cm. By tomographic method, heat transport outside from the 1/1 mode X-point during the sawtooth collapse is found. The other two cameras with a higher spatial resolution 1 cm are designed for monitoring local MHD activities respectively in plasma coremore » and boundary.« less

Two-dimensional vacuum ultraviolet images in different MHD events on the EAST tokamak

NASA Astrophysics Data System (ADS)

Zhijun, WANG; Xiang, GAO; Tingfeng, MING; Yumin, WANG; Fan, ZHOU; Feifei, LONG; Qing, ZHUANG; EAST Team

2018-02-01

A high-speed vacuum ultraviolet (VUV) imaging telescope system has been developed to measure the edge plasma emission (including the pedestal region) in the Experimental Advanced Superconducting Tokamak (EAST). The key optics of the high-speed VUV imaging system consists of three parts: an inverse Schwarzschild-type telescope, a micro-channel plate (MCP) and a visible imaging high-speed camera. The VUV imaging system has been operated routinely in the 2016 EAST experiment campaign. The dynamics of the two-dimensional (2D) images of magnetohydrodynamic (MHD) instabilities, such as edge localized modes (ELMs), tearing-like modes and disruptions, have been observed using this system. The related VUV images are presented in this paper, and it indicates the VUV imaging system is a potential tool which can be applied successfully in various plasma conditions.

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

US ITER Moving Forward

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

Sauthoff, Ned; Reiersen, Wayne; Berry, Jan

2013-09-12

US ITER Project Manager Ned Sauthoff, joined by Wayne Reiersen, Team Leader Magnet Systems, and Jan Berry, Team Leader Tokamak Cooling System, discuss the U.S.'s role in the ITER international collaboration.

US ITER Moving Forward

ScienceCinema

Sauthoff, Ned; Reiersen, Wayne; Berry, Jan

2017-12-12

US ITER Project Manager Ned Sauthoff, joined by Wayne Reiersen, Team Leader Magnet Systems, and Jan Berry, Team Leader Tokamak Cooling System, discuss the U.S.'s role in the ITER international collaboration.

High resolution main-ion charge exchange spectroscopy in the DIII-D H-mode pedestal

DOE PAGES

Grierson, B. A.; Burrell, K. H.; Chrystal, C.; ...

2016-09-12

A new high spatial resolution main-ion (deuterium) charge-exchange spectroscopy system covering the tokamak boundary region has been installed on the DIII-D tokamak. Sixteen new edge main-ion charge-exchange recombination sightlines have been combined with nineteen impurity sightlines in a tangentially viewing geometry on the DIII-D midplane with an interleaving design that achieves 8 mm inter-channel radial resolution for detailed profiles of main-ion temperature, velocity, charge-exchange emission, and neutral beam emission. At the plasma boundary, we find a strong enhancement of the main-ion toroidal velocity that exceeds the impurity velocity by a factor of two. Furthermore, the unique combination of experimentally measuredmore » main-ion and impurity profiles provides a powerful quasi-neutrality constraint for reconstruction of tokamak H-mode pedestals.« less

Tangential System of Thomson Scattering for Tokamak T-15

NASA Astrophysics Data System (ADS)

Asadulin, G. M.; Bel'bas, I. S.; Gorshkov, A. V.

2017-12-01

Two systems of Thomson scattering diagnostics, with vertical and tangential probing, are used in the D-shaped plasma cross section in tokamak T-15. The tangential system allows measuring plasma temperature and density profiles along the major radius of the tokamak. This paper presents the tangential system project. The system is based on a Nd:YAG laser with wavelength of 1064 nm, pulse energy of 3 J, pulse duration of 10 ns, and repetition rate of 100 Hz. The chosen geometry allows collecting light from ten uniformly spaced points. Optimization of the registration system has been accomplished. The collected light will be transmitted through an optical fiber bundle with diameter of 3 mm and quartz fibers (numerical aperture is 0.22). Six-channel polychromators based on high-contrast interference filters have been chosen as spectral equipment. The radiation will be registered by avalanche photodiodes. The technique of electron temperature and density measurement is described, and estimation of its accuracy is carried out. The proposed system allows measuring the electron temperature with accuracy not worse than 10% within the range of 50 eV to 10 keV on the pinch edge over the internal contour, from 20 eV to 9 keV in the plasma central region, and from 2 eV to 400 eV on the pinch edge over the outer contour. The estimation is made for electron density of not less than 2.6 × 1013 cm-3.

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.

Extending SIESTA capabilities: removing field-periodic and stellarator symmetric limitations

NASA Astrophysics Data System (ADS)

Cook, C. R.; Hirshman, S. P.; Sanchez, R.; Anderson, D. T.

2011-10-01

SIESTA is a three-dimensional magnetohydrodynamics equilibrium code capable of resolving magnetic islands in toroidal plasma confinement devices. Currently SIESTA assumes that plasma perturbations, and thus also magnetic islands, are field-periodic. This limitation is being removed from the code by allowing the displacement toroidal mode number to not be restricted to multiples of the number of field periods. Extending SIESTA in this manner will allow larger, lower-order resonant islands to form in devices such as CTH. An example of a non-field-periodic perturbation in CTH will be demonstrated. Currently the code also operates in a stellarator-symmetric fashion in which an ``up-down'' symmetry is present at some toroidal angle. Nearly all of the current tokamaks (and ITER in the future) operate with a divertor and as such do not possess stellarator symmetry. Removal of this symmetry restriction requires including both sine and cosine terms in the Fourier expansion for the geometry of the device and the fields contained within. The current status of this extension of the code will be discussed, along with the method of implementation. U.S. DOE Contract No. DE-AC05-00OR22725 with UT-Battelle, LLC.

Electron cyclotron heating/current-drive system using high power tubes for QUEST spherical tokamak

NASA Astrophysics Data System (ADS)

Onchi, Takumi; Idei, H.; Hasegawa, M.; Nagata, T.; Kuroda, K.; Hanada, K.; Kariya, T.; Kubo, S.; Tsujimura, T. I.; Kobayashi, S.; Quest Team

2017-10-01

Electron cyclotron heating (ECH) is the primary method to ramp up plasma current non-inductively in QUEST spherical tokamak. A 28 GHz gyrotron is employed for short pulses, where the radio frequency (RF) power is about 300 kW. Current ramp-up efficiency of 0.5 A/W has been obtained with focused beam of the second harmonic X-mode. A quasi-optical polarizer unit has been newly installed to avoid arcing events. For steady-state tokamak operation, 8.56 GHz klystron with power of 200 kW is used as the CW-RF source. The high voltage power supply (54 kV/13 A) for the klystron has been built recently, and initial bench test of the CW-ECH system is starting. The array of insulated-gate bipolar transistor works to quickly cut off the input power for protecting the klystron. This work is supported by JSPS KAKENHI (15H04231), NIFS Collaboration Research program (NIFS13KUTR085, NIFS17KUTR128), and through MEXT funding for young scientists associated with active promotion of national university reforms.

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.

Overview of RWM Stabilization and Other Experiments With New Internal Coils in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Jackson, G. L.; Evans, T. E.; La Haye, R. J.; Kellman, A. G.; Schaffer, M. J.; Scoville, J. T.; Strait, E. J.; Szymanski, D. D.; Bialek, J.; Garofalo, A. M.; Navratil, G. A.; Reimerdes, H.; Edgell, D. H.; Okabayashi, M.; Hatcher, R.

2003-10-01

A set of 12 single-turn internal coils (I-coils) has been installed and operated in the DIII-D tokamak. The primary purpose of these coils (A_coil = 1.1 m^2, I ≤,7 kA, d_wall = 1.47 cm) is to improve stabilization of the n=1 resistive wall mode (RWM), compared to the existing external C-coil set, especially for high βN advanced tokamak discharges in low toroidal rotation plasmas. The versatility of the I-coil set and its associated power systems allow for a variety of experiments: fast feedback stabilization of RWMs, dc error field correction, edge stochastic fields, n=1,2, or 3 toroidal magnetic braking, and MHD spectroscopy (0-60 Hz). The resonant field amplification from an applied n=1 field was found to be completely suppressed, demonstrating successfully the controllability with the new system. With the I-coils, the high βN regime (above the no wall limit) has been explored both with RWM feedback and with dynamic error field correction. Experiments on edge ergodization will also be discussed.

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

Liu, J. X., E-mail: jsliu9@berkeley.edu; Milbourne, T.; Bitter, M.

The implementation of advanced electron cyclotron emission imaging (ECEI) systems on tokamak experiments has revolutionized the diagnosis of magnetohydrodynamic (MHD) activities and improved our understanding of instabilities, which lead to disruptions. It is therefore desirable to have an ECEI system on the ITER tokamak. However, the large size of optical components in presently used ECEI systems have, up to now, precluded the implementation of an ECEI system on ITER. This paper describes a new optical ECEI concept that employs a single spherical mirror as the only optical component and exploits the astigmatism of such a mirror to produce an imagemore » with one-dimensional spatial resolution on the detector. Since this alternative approach would only require a thin slit as the viewing port to the plasma, it would make the implementation of an ECEI system on ITER feasible. The results obtained from proof-of-principle experiments with a 125 GHz microwave system are presented.« less

Neutral recycling effects on ITG turbulence

DOE PAGES

Stotler, D. P.; Lang, J.; Chang, C. S.; ...

2017-07-04

Here, the effects of recycled neutral atoms on tokamak ion temperature gradient (ITG) driven turbulence have been investigated in a steep edge pedestal, magnetic separatrix configuration, with the full-f edge gryokinetic code XGC1. An adiabatic electron model has been used; hence, the impacts of neutral particles and turbulence on the density gradient are not considered, nor are electromagnetic turbulence effects. The neutral atoms enhance the ITG turbulence, first, by increasing the ion temperature gradient in the pedestal via the cooling effects of charge exchange and, second, by a relative reduction in themore » $$E\\times B$$ shearing rate.« less

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.

Neutral recycling effects on ITG turbulence

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

Stotler, D. P.; Lang, J.; Chang, C. S.

Here, the effects of recycled neutral atoms on tokamak ion temperature gradient (ITG) driven turbulence have been investigated in a steep edge pedestal, magnetic separatrix configuration, with the full-f edge gryokinetic code XGC1. An adiabatic electron model has been used; hence, the impacts of neutral particles and turbulence on the density gradient are not considered, nor are electromagnetic turbulence effects. The neutral atoms enhance the ITG turbulence, first, by increasing the ion temperature gradient in the pedestal via the cooling effects of charge exchange and, second, by a relative reduction in themore » $$E\\times B$$ shearing rate.« less

X-ray Spectroscopy of E2 and M3 Transitions in Ni-like W

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

Clementson, J; Beiersdorfer, P; Gu, M F

2009-11-09

The electric quadrupole (E2) and magnetic octupole (M3) ground state transitions in Ni-like W{sup 46+} have been measured using high-resolution crystal spectroscopy at the Livermore electron beam ion trap facility. The lines fall in the soft x-ray region near 7.93 {angstrom} and were originally observed as an unresolved feature in tokamak plasmas. Using flat ADP and quartz crystals the wavelengths, intensities, and polarizations of the two lines have been measured for various electron beam energies and compared to intensity and polarization calculations performed using the Flexible Atomic Code (FAC).

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.

Spatial calibration of a tokamak neutral beam diagnostic using in situ neutral beam emission

DOE PAGES

Chrystal, Colin; Burrell, Keith H.; Grierson, Brian A.; ...

2015-10-20

Neutral beam injection is used in tokamaks to heat, apply torque, drive non-inductive current, and diagnose plasmas. Neutral beam diagnostics need accurate spatial calibrations to benefit from the measurement localization provided by the neutral beam. A new technique has been developed that uses in-situ measurements of neutral beam emission to determine the spatial location of the beam and the associated diagnostic views. This technique was developed to improve the charge exchange recombination diagnostic (CER) at the DIII-D tokamak and uses measurements of the Doppler shift and Stark splitting of neutral beam emission made by that diagnostic. These measurements contain informationmore » about the geometric relation between the diagnostic views and the neutral beams when they are injecting power. This information is combined with standard spatial calibration measurements to create an integrated spatial calibration that provides a more complete description of the neutral beam-CER system. The integrated spatial calibration results are very similar to the standard calibration results and derived quantities from CER measurements are unchanged within their measurement errors. Lastly, the methods developed to perform the integrated spatial calibration could be useful for tokamaks with limited physical access.« less

Development of frequency modulation reflectometer for Korea Superconducting Tokamak Advanced Research tokamak

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

Seo, Seong-Heon; Wi, H. M.; Lee, W. R.

2013-08-15

Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6–54 GHz), V band (48–72 GHz), and W band (72–108 GHz) to measure the density up to 7 × 10{sup 19} m{sup −3} when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank ofmore » low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.« less

Radial force on the vacuum chamber wall during thermal quench in tokamaks

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

Pustovitov, V. D., E-mail: pustovitov-vd@nrcki.ru

The radial force balance during a thermal quench in tokamaks is analyzed. As a rule, the duration τ{sub tp} of such events is much shorter than the resistive time τ{sub w} of the vacuum chamber wall. Therefore, the perturbations of the magnetic field B produced by the evolving plasma cannot penetrate the wall, which makes different the magnetic pressures on its inner and outer sides. The goal of this work is the analytical estimation of the resulting integral radial force on the wall. The plasma is considered axially symmetric; for the description of radial forces on the wall, the resultsmore » of V.D. Shafranov’s classical work [J. Nucl. Energy C 5, 251 (1963)] are used. Developed for tokamaks, the standard equilibrium theory considers three interacting systems: plasma, poloidal field coils, and toroidal field coils. Here, the wall is additionally incorporated with currents driven by ∂B/∂t≠0 accompanying the fast loss of the plasma thermal energy. It is shown that they essentially affect the force redistribution, thereby leading to large loads on the wall. The estimates prove that these loads have to be accounted for in the disruptive scenarios in large tokamaks.« less

Spatial calibration of a tokamak neutral beam diagnostic using in situ neutral beam emission

NASA Astrophysics Data System (ADS)

Chrystal, C.; Burrell, K. H.; Grierson, B. A.; Pace, D. C.

2015-10-01

Neutral beam injection is used in tokamaks to heat, apply torque, drive non-inductive current, and diagnose plasmas. Neutral beam diagnostics need accurate spatial calibrations to benefit from the measurement localization provided by the neutral beam. A new technique has been developed that uses in situ measurements of neutral beam emission to determine the spatial location of the beam and the associated diagnostic views. This technique was developed to improve the charge exchange recombination (CER) diagnostic at the DIII-D tokamak and uses measurements of the Doppler shift and Stark splitting of neutral beam emission made by that diagnostic. These measurements contain information about the geometric relation between the diagnostic views and the neutral beams when they are injecting power. This information is combined with standard spatial calibration measurements to create an integrated spatial calibration that provides a more complete description of the neutral beam-CER system. The integrated spatial calibration results are very similar to the standard calibration results and derived quantities from CER measurements are unchanged within their measurement errors. The methods developed to perform the integrated spatial calibration could be useful for tokamaks with limited physical access.

Impact of helical boundary conditions in MHD modeling of RFP and tokamak plasmas

NASA Astrophysics Data System (ADS)

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

2012-10-01

Helical boundary conditions imposed by the active control system of the RFX-mod device provide a handle to govern the plasma dynamics in both RFP and Ohmic tokamak discharges [1]. By applying an edge radial magnetic field with proper helicity, it is possible to increase the persistence of the spontaneous helical RFP states at high current,and to stimulate them also at low current or high density. Helical BCs even allow to access helical states with different helicity than the spontaneous one [2]. In Ohmic tokamak operation at q(a)<2, the presence of the 2/1 RWM reduces the sawtoothing activity of the 1/1 internal kink, which takes a stationary snake-like character instead. Many of these features are qualitatively reproduced in 3D nonlinear MHD modeling. We study the impact of helical BCs on the MHD dynamics in both RFP and tokamak with two successfully benchmarked numerical tools, SpeCyl and PIXIE3D [3]. We recover the bifurcation from a sawtooth to a snake solution when imposing a 2/1 BC in the tokamak case and we interpret this as a toroidal/nonlinear coupling effect. We show that the bifurcation is more easily stimulated with a 1/1 BC.[4pt] [1] P. Piovesan, invited talk this meeting[0pt] [2] M. Veranda et al EPS-ICPP Conference (2012) P4.004[0pt] [3] D. Bonfiglio et al Phys. Plasmas (2010)

The Multiple Gyrotron System on the DIII-D Tokamak

DOE PAGES

Lohr, J.; Anderson, J.; Brambila, R.; ...

2015-08-28

A major component of the versatile heating systems on the DIII-D tokamak is the gyrotron complex. This system routinely operates at 110 GHz with 4.7 MW generated rf power for electron cyclotron heating and current drive. The complex is being upgraded with the addition of new depressed collector potential gyrotrons operating at 117.5 GHz and generating rf power in excess of 1.0 MW each. The long term upgrade plan calls for 10 gyrotrons at the higher frequency being phased in as resources permit, for an injected power near 10 MW. This article presents a summary of the current status ofmore » the DIII-D gyrotron complex, its performance, individual components, testing procedures, operational parameters, plans, and a brief summary of the experiments for which the system is currently being used.« less

Pellet injection research on the HT-6M and HT-7 tokamaks

NASA Astrophysics Data System (ADS)

Yang, Yu; Bao, Yi; Li, Jiangang; Gu, Xuemao; He, Yexi

1999-11-01

A multishot in situ pellet injection system has been constructed in the Institute of Plasma Physics. Single- and multi-pellet injection experiments were performed on the HT-6M and superconducting HT-7 tokamaks. The system proved to be convenient and reliable to operate. Pellets were fired into ohmically and LHCD and ICRF heated plasmas. Single pellet injection in ohmic discharge was found to increase the central density of HT-7 by about one half, while two pellet injection increased the central density in a step-like fashion by one half with each shot. Peaking of the electron density profile and a hollow electron temperature profile were obtained.

Bench Test of the Vibration Compensation Interferometer for EAST Tokamak

NASA Astrophysics Data System (ADS)

Li, Gongshun; Yang, Yao; Liu, Haiqing; Jie, Yinxian; Zou, Zhiyong; Wang, Zhengxing; Zeng, Long; Wei, Xuechao; Li, Weiming; Lan, Ting; Zhu, Xiang; Liu, Yukai; Gao, Xiang

2016-02-01

A visible laser-based vibration compensation interferometer has recently been designed for the EAST tokamak and the bench test has been finished. The system was optimized for its installation on EAST. The value of the final optical power before the detectors without plasma has been calculated from the component bench test result, which is quite close to the measured value. A nanometer level displacement (of the order of the laser's wavelength) has been clearly measured by a modulation of piezoelectric ceramic unit, proving the system's capability. supported by the National Magnetic Confinement Fusion Program of China (Nos. 2014GB106002, 2014GB106003, 2014GB106004) and National Natural Science Foundation of China (Nos. 11105184, 11375237, 11505238)

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

Redi, M.H.; Mynick, H.E.; Suewattana, M.

Hamiltonian coordinate, guiding-center code calculations of the confinement of suprathermal ions in quasi-axisymmetric stellarator (QAS) designs have been carried out to evaluate the attractiveness of compact configurations which are optimized for ballooning stability. A new stellarator particle-following code is used to predict ion loss rates and particle confinement for thermal and neutral beam ions in a small experiment with R = 145 cm, B = 1-2 T and for alpha particles in a reactor-size device. In contrast to tokamaks, it is found that high edge poloidal flux has limited value in improving ion confinement in QAS, since collisional pitch-angle scatteringmore » drives ions into ripple wells and stochastic field regions, where they are quickly lost. The necessity for reduced stellarator ripple fields is emphasized. The high neutral beam ion loss predicted for these configurations suggests that more interesting physics could be explored with an experiment of less constrained size and magnetic field geometry.« less

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.

Stability properties and fast ion confinement of hybrid tokamak plasma configurations

NASA Astrophysics Data System (ADS)

Graves, J. P.; Brunetti, D.; Pfefferle, D.; Faustin, J. M. P.; Cooper, W. A.; Kleiner, A.; Lanthaler, S.; Patten, H. W.; Raghunathan, M.

2015-11-01

In hybrid scenarios with flat q just above unity, extremely fast growing tearing modes are born from toroidal sidebands of the near resonant ideal internal kink mode. New scalings of the growth rate with the magnetic Reynolds number arise from two fluid effects and sheared toroidal flow. Non-linear saturated 1/1 dominant modes obtained from initial value stability calculation agree with the amplitude of the 1/1 component of a 3D VMEC equilibrium calculation. Viable and realistic equilibrium representation of such internal kink modes allow fast ion studies to be accurately established. Calculations of MAST neutral beam ion distributions using the VENUS-LEVIS code show very good agreement of observed impaired core fast ion confinement when long lived modes occur. The 3D ICRH code SCENIC also enables the establishment of minority RF distributions in hybrid plasmas susceptible to saturated near resonant internal kink modes.

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.

Automation of the guiding center expansion

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

Burby, J. W.; Squire, J.; Qin, H.

2013-07-15

We report on the use of the recently developed Mathematica package VEST (Vector Einstein Summation Tools) to automatically derive the guiding center transformation. Our Mathematica code employs a recursive procedure to derive the transformation order-by-order. This procedure has several novel features. (1) It is designed to allow the user to easily explore the guiding center transformation's numerous non-unique forms or representations. (2) The procedure proceeds entirely in cartesian position and velocity coordinates, thereby producing manifestly gyrogauge invariant results; the commonly used perpendicular unit vector fields e{sub 1},e{sub 2} are never even introduced. (3) It is easy to apply in themore » derivation of higher-order contributions to the guiding center transformation without fear of human error. Our code therefore stands as a useful tool for exploring subtle issues related to the physics of toroidal momentum conservation in tokamaks.« less

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.

Remote network control plasma diagnostic system for Tokamak T-10

NASA Astrophysics Data System (ADS)

Troynov, V. I.; Zimin, A. M.; Krupin, V. A.; Notkin, G. E.; Nurgaliev, M. R.

2016-09-01

The parameters of molecular plasma in closed magnetic trap is studied in this paper. Using the system of molecular diagnostics, which was designed by the authors on the «Tokamak T-10» facility, the radiation of hydrogen isotopes at the plasma edge is investigated. The scheme of optical radiation registration within visible spectrum is described. For visualization, identification and processing of registered molecular spectra a new software is developed using MatLab environment. The software also includes electronic atlas of electronic-vibrational-rotational transitions for molecules of protium and deuterium. To register radiation from limiter cross-section a network control system is designed using the means of the Internet/Intranet. Remote control system diagram and methods are given. The examples of web-interfaces for working out equipment control scenarios and viewing of results are provided. After test run in Intranet, the remote diagnostic system will be accessible through Internet.

Comparative analysis of the possibility of applying low-melting metals with the capillary-porous system in tokamak conditions

NASA Astrophysics Data System (ADS)

Lyublinski, I. E.; Vertkov, A. V.; Semenov, V. V.

2016-12-01

The use of capillary-porous systems (CPSs) with liquid Li, Ga, and Sn is considered as an alternative for solving the problem of creating plasma-facing elements (PFEs) of the fusion neutron source (FNS) and the DEMO-type reactor. The main advantages of CPSs with liquid metal compared with hard materials are their stability with respect to the degradation of properties in tokamak conditions and capability of surface self-restoration. The evaluation of applicability of liquid metals is performed on the basis of the analysis of their physical and chemical properties, the interaction with the tokamak plasma, and constructive and process features of in-vessel elements with CPSs implementing the application of these metals in a tokamak. It is shown that the upper limit of the PFE working temperature for all low-melting metals under consideration lies in the range of 550-600°C. The decisive factor for PFEs with Li is the limitation on the admissible atomic flux into plasma, while for those with Ga and Sn it is the corrosion resistance of construction materials. The upper limit of thermal loads in the steady-state operating mode for the considered promising PFE design with the use of Li, Ga, and Sn is close to 18-20 MW/m2. It is seen from the analysis that the use of metals with a low equilibrium vapor pressure of (Ga, Sn) gives no gain in extension of the region of admissible working temperatures of PFEs. However, with respect to the totality of properties, the possibility of implementing the self-restoration and stabilization effect of the liquid surface, the influence on the plasma discharge parameters, and the ability to protect the PFE surface in conditions of plasma perturbations and disruption, lithium is the most attractive liquid metal to create CPS-based PFEs for the tokamak.

Comparative analysis of the possibility of applying low-melting metals with the capillary-porous system in tokamak conditions

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

Lyublinski, I. E., E-mail: lyublinski@yandex.ru; Vertkov, A. V., E-mail: avertkov@yandex.ru; Semenov, V. V., E-mail: darkfenix2006@mail.ru

2016-12-15

The use of capillary-porous systems (CPSs) with liquid Li, Ga, and Sn is considered as an alternative for solving the problem of creating plasma-facing elements (PFEs) of the fusion neutron source (FNS) and the DEMO-type reactor. The main advantages of CPSs with liquid metal compared with hard materials are their stability with respect to the degradation of properties in tokamak conditions and capability of surface self-restoration. The evaluation of applicability of liquid metals is performed on the basis of the analysis of their physical and chemical properties, the interaction with the tokamak plasma, and constructive and process features of in-vesselmore » elements with CPSs implementing the application of these metals in a tokamak. It is shown that the upper limit of the PFE working temperature for all low-melting metals under consideration lies in the range of 550–600°Ð¡. The decisive factor for PFEs with Li is the limitation on the admissible atomic flux into plasma, while for those with Ga and Sn it is the corrosion resistance of construction materials. The upper limit of thermal loads in the steady-state operating mode for the considered promising PFE design with the use of Li, Ga, and Sn is close to 18–20 MW/m{sup 2}. It is seen from the analysis that the use of metals with a low equilibrium vapor pressure of (Ga, Sn) gives no gain in extension of the region of admissible working temperatures of PFEs. However, with respect to the totality of properties, the possibility of implementing the self-restoration and stabilization effect of the liquid surface, the influence on the plasma discharge parameters, and the ability to protect the PFE surface in conditions of plasma perturbations and disruption, lithium is the most attractive liquid metal to create CPS-based PFEs for the tokamak.« less

RF assisted Glow Discharge Condition experiment for SST-1 Tokamak

NASA Astrophysics Data System (ADS)

Raval, Dilip; Khan, Ziauddin; George, Siju; Dhanani, Kalpeshkumar R.; Paravastu, Yuvakiran; Semwal, Pratibha; Thankey, Prashant; Shoaib Khan, Mohammad; Kakati, Bharat; Pradhan, Subrata

2017-04-01

Impurity control reduces the radiation loss from plasma and hence enhances the plasma operation. Oxygen and water vapors are the most common impurities in tokamak devices. Water vapour can be reduced with extensive baking while in order to have a significant reduction in oxygen it is necessary to use glow discharge condition (GDC). RF assisted glow discharge cleaning system will be implemented to remove low z impurities at PFC installed SST-1 vacuum vessel. A RF assisted Glow discharge conditioning is studied at laboratory to find the optimum operating parameters in a view to implement at SST-1 tokamak. Helium is used as a fuel gas in the present experiment. It is observed that the ultimate impurity level is reduced significantly below to the accepted level for plasma operation after RF assisted GDC. The experimental findings of RF assisted Glow discharge conditioning is discussed in details in this paper.

Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas.

PubMed

Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M

2012-01-10

Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas.

Advances in Dust Detection and Removal for Tokamaks

NASA Astrophysics Data System (ADS)

Campos, A.; Skinner, C. H.; Roquemore, A. L.; Leisure, J. O. V.; Wagner, S.

2008-11-01

Dust diagnostics and removal techniques are vital for the safe operation of next step fusion devices such as ITER. An electrostatic dust detector[1] developed in the laboratory is being applied to NSTX. In the tokamak environment, large particles or fibres can fall on the grid potentially causing a permanent short. We report on the development of a gas puff system that uses helium to clear such particles from the detector. Experiments with varying nozzle designs, backing pressures, puff durations, and exit flow orientations have obtained an optimal configuration that effectively removes particles from a 25 cm^2 area. Dust removal from next step tokamaks will be required to meet regulatory dust limits. A tripolar grid of fine interdigitated traces has been designed that generates an electrostatic travelling wave for conveying dust particles to a ``drain.'' First trials have shown particle motion in optical microscope images. [1] C. H. Skinner et al., J. Nucl. Mater., 376 (2008) 29.

Workshop on High Power ICH Antenna Designs for High Density Tokamaks

NASA Astrophysics Data System (ADS)

Aamodt, R. E.

1990-02-01

A workshop in high power ICH antenna designs for high density tokamaks was held to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of RF auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made.

Estimation of Electron Temperature on Glass Spherical Tokamak (GLAST)

NASA Astrophysics Data System (ADS)

Hussain, S.; Sadiq, M.; Shah, S. I. W.; GLAST Team

2015-03-01

Glass Spherical Tokamak (GLAST) is a small spherical tokamak indigenously developed in Pakistan with an insulating vacuum vessel. A commercially available 2.45 GHz magnetron is used as pre-ionization source for plasma current startup. Different diagnostic systems like Rogowski coils, magnetic probes, flux loops, Langmuir probe, fast imaging and emission spectroscopy are installed on the device. The plasma temperature inside of GLAST, at the time of maxima of plasma current, is estimated by taking into account the Spitzer resistivity calculations with some experimentally determined plasma parameters. The plasma resistance is calculated by using Ohm's law with plasma current and loop voltage as experimentally determined inputs. The plasma resistivity is then determined by using length and area of the plasma column. Finally, the average plasma electron temperature is predicted to be 12.65eV for taking neon (Ne) as a working gas.

The DIII-D Plasma Control System as a Scientific Research Tool

NASA Astrophysics Data System (ADS)

Hyatt, A. W.; Ferron, J. R.; Humphreys, D. A.; Leuer, J. A.; Walker, M. L.; Welander, A. S.

2006-10-01

The digital plasma control system (PCS) is an essential element of the DIII-D tokamak as a scientific research instrument, providing experimenters with real-time measurement and control of the plasma equilibrium, heating, current drive, transport, stability, and plasma-wall interactions. A wide range of sensors and actuators allow feedback control not only of global quantities such as discharge shape, plasma energy, and toroidal rotation, but also of non-axisymmetric magnetic fields and features of the internal profiles of temperature and current density. These diverse capabilities of the PCS improve the effectiveness of tokamak operation and enable unique physics experiments. We will present an overview of the PCS and the systems it controls and interacts with, and show examples of various plasma parameters controlled by the PCS and its actuators.

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.