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  1. Partnership for Edge Physics Simulation

    A major goal of our participation in the Edge Physics Simulation project has been to contribute to the understanding of the self-organization of tokamak turbulence fluctuations resulting in the formation of a staircase structure in the ion temperature. A second important goal is to demonstrate how small scale turbulence in plasmas self-organizes with dynamically driven quasi-stationary flow shear. These goals have been accomplished through the analyses of the statistical properties of XGC1 flux driven Gyrokinetic electrostatic ion temperature gradient (ITG) turbulence simulation data in which neutrals are included. The ITG turbulence data, and in particular fluctuation data, were obtained frommore » a massively parallel flux-driven gyrokinetic full-f particle-in-cell simulation of a DIII-D like equilibrium. Below some the findings are summarized. It was observed that the emergence of staircase structure is related to the variations in the normalized temperature gradient length (R/LT) and the poloidal flow shear. Average turbulence intensity is found to be large in the vicinity of minima in R/LTi, where ITG growth is expected to be lower. The distributions of the occurrences of potential fluctuation are found to be Gaussian away from the staircase-step locations, but they are found to be non-Gaussian in the vicinity of staircase-step locations. The results of analytically derived expressions for the distribution of the occurrences of turbulence intensity and intensity flux were compared with the corresponding quantities computed using XGC1 simulation data and good agreement is found. The derived expressions predicts inward and outward propagation of turbulence intensity flux in an intermittent fashion. The outward propagation of turbulence intensity flux occurs at staircase-step locations and is related to the change in poloidal flow velocity shear and to the change in the ion temperature gradient. The standard deviation, skewness and kurtosis for turbulence quantities were computed and found to be large in the vicinity of the staircase-step structures. Large values of skewness and kurtosis can be explained by a temporary opening and closing of the structure which allows turbulence intensity events to propagate. The staircase patterns may reduce the ion heat transport and a manipulation of these patterns may be used to optimize heat transport in tokamaks. An additional objective of the research in support of the Edge Physics Simulation initiative has been to improve the understanding of scrape-off layer thermal transport. In planning experiments and designing future tokamaks, it is important to understand the physical effects that contribute to divertor heat-load fluxes. The research accomplished will contribute to developing new models for the scrape-off layer region. The XGC0 code was used to compute the heat fluxes and the heat-load width in the outer divertor plates of C-Mod and DIII-D tokamaks. It was observed that the width of the XGC0 neoclassical heat-load was approximately inversely proportional to the total plasma current. Anomalous transport in the H-mode pedestal region of five Alcator C-Mod discharges, representing a collisionality scan, was analyzed. The understanding of anomalous transport in the pedestal region is important for the development of a comprehensive model for the H-mode pedestal slope. It was found that the electron thermal anomalous diffusivities at the pedestal top increase with the electron collisionality. This dependence can point to the DRIBM as the modes that drive the anomalous transport in the plasma edge of highly collisional discharges. The effects of plasma shaping on the H-mode pedestal structure was also investigated. The differences in the predicted H-mode pedestal width and height for the DIII-D discharges with different elongation and triangularities were discussed. For the discharges with higher elongation, it was found that the gradients of the plasma profiles in the H-mode pedestal reach semi-steady states. In these simulations, the pedestal slowly continued to evolve to higher pedestal pressures and bootstrap currents until the peeling ballooning stability conditions were satisfied. The discharges with lower elongation do not reach the semi-steady state, and ELM crashes were triggered at earlier times. The plasma elongation was found to have a stronger stabilizing effect than the plasma triangularity. For the discharges with lower elongation and lower triangularity, the ELM frequency was large, and the H-mode pedestal evolves rapidly. It was found that the temperature of neutrals in the scrape-off-layer region can affect the dynamics of the H-mode pedestal buildup. However, the final pedestal profiles were nearly independent of the neutral temperature. The elongation and triangularity affected the pedestal widths of plasma density and electron temperature profiles differently. This study illustrated a new mechanism for controlling the pedestal bootstrap current and the pedestal stability.« less
  2. Investigation of ELM [edge localized mode] Dynamics with the Resonant Magnetic Perturbation Effects

    Topics covered are: anomalous transport and E x B flow shear effects in the H-mode pedestal; RMP (resonant magnetic perturbation) effects in NSTX discharges; development of a scaling of H-mode pedestal in tokamak plasmas with type I ELMs (edge localized modes); and divertor heat load studies.
  3. Final Technical Report for Center for Plasma Edge Simulation Research

    The CPES research carried out by the Lehigh fusion group has sought to satisfy the evolving requirements of the CPES project. Overall, the Lehigh group has focused on verification and validation of the codes developed and/or integrated in the CPES project. Consequently, contacts and interaction with experimentalists have been maintained during the course of the project. Prof. Arnold Kritz, the leader of the Lehigh Fusion Group, has participated in the executive management of the CPES project. The code development and simulation studies carried out by the Lehigh fusion group are described in more detail in the sections below.
  4. Fusion Energy Sciences Exascale Requirements Review. An Office of Science review sponsored jointly by Advanced Scientific Computing Research and Fusion Energy Sciences, January 27-29, 2016, Gaithersburg, Maryland

    The additional computing power offered by the planned exascale facilities could be transformational across the spectrum of plasma and fusion research — provided that the new architectures can be efficiently applied to our problem space. The collaboration that will be required to succeed should be viewed as an opportunity to identify and exploit cross-disciplinary synergies. To assess the opportunities and requirements as part of the development of an overall strategy for computing in the exascale era, the Exascale Requirements Review meeting of the Fusion Energy Sciences (FES) community was convened January 27–29, 2016, with participation from a broad range ofmore » fusion and plasma scientists, specialists in applied mathematics and computer science, and representatives from the U.S. Department of Energy (DOE) and its major computing facilities. This report is a summary of that meeting and the preparatory activities for it and includes a wealth of detail to support the findings. Technical opportunities, requirements, and challenges are detailed in this report (and in the recent report on the Workshop on Integrated Simulation). Science applications are described, along with mathematical and computational enabling technologies. Also see http://exascaleage.org/fes/ for more information.« less
  5. Model for current drive stabilization of neoclassical tearing modes

    A model derivation is presented for the effect of current drive on the saturated width of magnetic islands driven by the neoclassical tearing mode instability in axisymmetric plasmas. The derivation is carried out for continuous current drive as well as for modulated current that is driven at the same angle as the island O-point. The results of the derivation are implemented in a revision of the ISLAND module to compute saturated magnetic island widths. It is found that the greatest stabilizing effect of both modulated and continuous current drive on the island width is achieved when current is driven atmore » the island center. In addition, narrow current drive is more effective at stabilizing the magnetic islands than wide current drive, for which more current falls outside the island. When modulated and continuous current drives are compared for equal total driven current, the modulated current is shown to be more effective, particularly as the offset from the island center increases.« less
  6. Predictive Simulations of ITER Including Neutral Beam Driven Toroidal Rotation

    Predictive simulations of ITER [R. Aymar et al., Plasma Phys. Control. Fusion 44, 519 2002] discharges are carried out for the 15 MA high confinement mode (H-mode) scenario using PTRANSP, the predictive version of the TRANSP code. The thermal and toroidal momentum transport equations are evolved using turbulent and neoclassical transport models. A predictive model is used to compute the temperature and width of the H-mode pedestal. The ITER simulations are carried out for neutral beam injection (NBI) heated plasmas, for ion cyclotron resonant frequency (ICRF) heated plasmas, and for plasmas heated with a mix of NBI and ICRF. Itmore » is shown that neutral beam injection drives toroidal rotation that improves the confinement and fusion power production in ITER. The scaling of fusion power with respect to the input power and to the pedestal temperature is studied. It is observed that, in simulations carried out using the momentum transport diffusivity computed using the GLF23 model [R.Waltz et al., Phys. Plasmas 4, 2482 (1997)], the fusion power increases with increasing injected beam power and central rotation frequency. It is found that the ITER target fusion power of 500 MW is produced with 20 MW of NBI power when the pedesta temperature is 3.5 keV. 2008 American Institute of Physics. [DOI: 10.1063/1.2931037]« less
  7. Comparison of low confinement mode transport simulations using the mixed Bohm/gyro-Bohm and the Multi-Mode-95 transport model

    Predictive transport simulations using the mixed Bohm/gyro-Bohm (JET) transport model [M. Erba , Plasma Phys. Controlled Fusion 39, 261 (1997)] are compared with simulations using the Multi-Mode-95 (MMM95) transport model [G. Bateman , Phys. Plasmas 5, 1793 (1998)]. Temperature and density profiles from these simulations are compared with experimental data for 13 low confinement mode (L-mode) discharges from the Doublet III-D Tokamak (DIII-D) [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)] and the Tokamak Fusion Test Reactor (TFTR) [D. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)]. The selected discharges include systematic scans overmore » gyro-radius, plasma power, current, and density. It is found that simulations using the two models match experimental data equally well, in spite of the fact that the JET model has predominantly Bohm scaling (proportional to gyro-radius) while the MMM95 model has a purely gyro-Bohm scaling (proportional to gyro-radius squared).« less
  8. Integrated simulations of saturated neoclassical tearing modes in DIII-D, Joint European Torus, and ITER plasmas

    A revised version of the ISLAND module [C. N. Nguyen et al., Phys. Plasmas 11, 3604 (2004)] is used in the BALDUR code [C. E. Singer et al., Comput. Phys. Commun. 49, 275 (1988)] to carry out integrated modeling simulations of DIII-D [J. Luxon, Nucl. Fusion 42, 614 (2002)], Joint European Torus (JET) [P. H. Rebut et al., Nucl. Fusion 25, 1011 (1985)], and ITER [R. Aymar et al., Plasma Phys. Control. Fusion 44, 519 (2002)] tokamak discharges in order to investigate the adverse effects of multiple saturated magnetic islands driven by neoclassical tearing modes (NTMs). Simulations are carried outmore » with a predictive model for the temperature and density pedestal at the edge of the high confinement mode (H-mode) plasma and with core transport described using the Multi-Mode model. The ISLAND module, which is used to compute magnetic island widths, includes the effects of an arbitrary aspect ratio and plasma cross sectional shape, the effect of the neoclassical bootstrap current, and the effect of the distortion in the shape of each magnetic island caused by the radial variation of the perturbed magnetic field. Radial transport is enhanced across the width of each magnetic island within the BALDUR integrated modeling simulations in order to produce a self-consistent local flattening of the plasma profiles. It is found that the main consequence of the NTM magnetic islands is a decrease in the central plasma temperature and total energy. For the DIII-D and JET discharges, it is found that inclusion of the NTMs typically results in a decrease in total energy of the order of 15%. In simulations of ITER, it is found that the saturated magnetic island widths normalized by the plasma minor radius, for the lowest order individual tearing modes, are approximately 24% for the 2/1 mode and 12% for the 3/2 mode. As a result, the ratio of ITER fusion power to heating power (fusion Q) is reduced from Q=10.6 in simulations with no NTM islands to Q=2.6 in simulations with fully saturated NTM islands.« less
  9. Predictive simulations of ITER including neutral beam driven toroidal rotation

    Predictive simulations of ITER [R. Aymar et al., Plasma Phys. Control. Fusion 44, 519 (2002)], discharges are carried out for the 15 MA high confinement mode (H-mode) scenario using PTRANSP, the predictive version of the TRANSP code. The thermal and toroidal momentum transport equations are evolved using turbulent and neoclassical transport models. A predictive model is used to compute the temperature and width of the H-mode pedestal. The ITER simulations are carried out for neutral beam injection (NBI) heated plasmas, for ion cyclotron resonant frequency (ICRF) heated plasmas, and for plasmas heated with a mix of NBI and ICRF. Itmore » is shown that neutral beam injection drives toroidal rotation that improves the confinement and fusion power production in ITER. The scaling of fusion power with respect to the input power and to the pedestal temperature is studied. It is observed that, in simulations carried out using the momentum transport diffusivity computed using the GLF23 model [R. Waltz et al., Phys. Plasmas 4, 2482 (1997)], the fusion power increases with increasing injected beam power and central rotation frequency. It is found that the ITER target fusion power of 500 MW is produced with 20 MW of NBI power when the pedestal temperature is 3.5 keV.« less
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