Showing papers by "Princeton Plasma Physics Laboratory published in 2011"

Compressing the incompressible with ISABELA: in-situ reduction of spatio-temporal data

Abstract: Modern large-scale scientific simulations running on HPC systems generate data in the order of terabytes during a single run. To lessen the I/O load during a simulation run, scientists are forced to capture data infrequently, thereby making data collection an inherently lossy process. Yet, lossless compression techniques are hardly suitable for scientific data due to its inherently random nature; for the applications used here, they offer less than 10% compression rate. They also impose significant overhead during decompression, making them unsuitable for data analysis and visualization that require repeated data access. To address this problem, we propose an effective method for In-situ Sort-And-B-spline Error-bounded Lossy Abatement (ISABELA) of scientific data that is widely regarded as effectively incompressible. With ISABELA, we apply a preconditioner to seemingly random and noisy data along spatial resolution to achieve an accurate fitting model that guarantees a ≥ 0.99 correlation with the original data. We further take advantage of temporal patterns in scientific data to compress data by ≈ 85%, while introducing only a negligible overhead on simulations in terms of runtime. ISABELA significantly outperforms existing lossy compression methods, such as Wavelet compression. Moreover, besides being a communication-free and scalable compression technique, ISABELA is an inherently local decompression method, namely it does not decode the entire data, making it attractive for random access.

Two-dimensional visualization of growth and burst of the edge-localized filaments in KSTAR H-mode plasmas.

Abstract: The filamentary nature and dynamics of edge-localized modes (ELMs) in the KSTAR high-confinement mode plasmas have been visualized in 2D via electron cyclotron emission imaging. The ELM filaments rotating with a net poloidal velocity are observed to evolve in three distinctive stages: initial linear growth, interim quasisteady state, and final crash. The crash is initiated by a narrow fingerlike perturbation growing radially from a poloidally elongated filament. The filament bursts through this finger, leading to fast and collective heat convection from the edge region into the scrape-off layer, i.e., ELM crash.

Electromagnetic transport from microtearing mode turbulence.

Abstract: This Letter presents non-linear gyrokinetic simulations of microtearing mode turbulence. The simulations include collisional and electromagnetic effects and use experimental parameters from a high beta discharge in the National Spherical Torus Experiment (NSTX). The predicted electron thermal transport is comparable to that given by experimental analysis, and it is dominated by the electromagnetic contribution of electrons free streaming along the resulting stochastic magnetic field line trajectories. Experimental values of flow shear can significantly reduce the predicted transport.

Advances towards QH-mode viability for ELM-stable operation in ITER

Abstract: The application of static, non-axisymmetric, nonresonant magnetic fields (NRMFs) to high beta DIII-D plasmas has allowed sustained operation with a quiescent H-mode (QH-mode) edge and both toroidal rotation and neutral beam injected torque near zero. Previous studies have shown that QH-mode operation can be accessed only if sufficient radial shear in the plasma flow is produced near the plasma edge. In past experiments, this flow shear was produced using neutral beam injection (NBI) to provide toroidal torque. In recent experiments, this torque was nearly completely replaced by the torque from applied NRMFs. The application of the NRMFs does not degrade the global energy confinement of the plasma. Conversely, the experiments show that the energy confinement quality increases with lower plasma rotation. Furthermore, the NRMF torque increases plasma resilience to locked modes at low rotation. These results open a path towards QH-mode utilization as an edge-localized mode (ELM)-stable H-mode in the self-heated burning plasma scenario, where toroidal momentum input from NBI may be small or absent.

Study on H-mode access at low density with lower hybrid current drive and lithium-wall coatings on the EAST superconducting tokamak

Abstract: The first high-confinement mode (H-mode) with type-III edge localized modes at an H factor of HIPB98(y,2) ∼ 1 has been obtained with about 1 MW lower hybrid wave power on the EAST superconducting tokamak. The first H-mode plasma appeared after wall conditioning by lithium (Li) evaporation before plasma breakdown and the real-time injection of fine Li powder into the plasma edge. The threshold power for H-mode access follows the international tokamak scaling even in the low density range and a threshold in density has been identified. With increasing accumulation of deposited Li the H-mode duration was gradually extended up to 3.6 s corresponding to ∼30 confinement times, limited only by currently attainable durations of the plasma current flat top. Finally, it was observed that neutral density near the lower X-point was progressively reduced by a factor of 4 with increasing Li accumulation, which is considered the main mechanism for the H-mode power threshold reduction by the Li wall coatings. (Some figures in this article are in colour only in the electronic version)

Fast Ion Induced Shearing of 2D Alfven Eigenmodes Measured by Electron Cyclotron Emission Imaging

Abstract: Two-dimensional images of electron temperature perturbations are obtained with electron cyclotron emission imaging (ECEI) on the DIII-D tokamak and compared to Alfven eigenmode structures obtained by numerical modeling using both ideal MHD and hybrid MHD-gyrofluid codes. While many features of the observations are found to be in excellent agreement with simulations using an ideal MHD code (NOVA), other characteristics distinctly reveal the influence of fast ions on the mode structures. These features are found to be well described by the nonperturbative hybrid MHD-gyrofluid model TAEFL.

Lithium?metal infused trenches (LiMIT) for heat removal in fusion devices

Abstract: Observation of liquid lithium flow in metal trenches has been made using a lithium–metal infused trench (LiMIT) tile and is reported here The flow is self-pumping and uses thermoelectric magnetohydrodynamics to remove heated lithium and replenish it at a lower temperature Flow velocities have been measured and compared with theoretical predictions

L-mode validation studies of gyrokinetic turbulence simulations via multiscale and multifield turbulence measurements on the DIII-D tokamak

Abstract: A series of carefully designed experiments on DIII-D have taken advantage of a broad set of turbulence and profile diagnostics to rigorously test gyrokinetic turbulence simulations. In this paper the goals, tools and experiments performed in these validation studies are reviewed and specific examples presented. It is found that predictions of transport and fluctuation levels in the mid-core region (0.4 < ρ < 0.75) are in better agreement with experiment than those in the outer region (ρ ≥ 0.75) where edge coupling effects may become increasingly important and multiscale simulations may also be necessary. Validation studies such as these are crucial in developing confidence in a first-principles based predictive capability for ITER.

Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

Abstract: A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

On velocity space interrogation regions of fast-ion collective Thomson scattering at ITER

Abstract: The collective Thomson scattering (CTS) diagnostic proposed for ITER is designed to measure projected 1D fast-ion velocity distribution functions at several spatial locations simultaneously. The frequency shift of scattered radiation and the scattering geometry place fast ions that caused the collective scattering in well-defined regions in velocity space, here dubbed interrogation regions. Since the CTS instrument measures entire spectra of scattered radiation, many different interrogation regions are probed simultaneously. We here give analytic expressions for weight functions describing the interrogation regions, and we show typical interrogation regions of the proposed ITER CTS system. The backscattering system with receivers on the low-field side is sensitive to fast ions with pitch |p| = |v∥/v| 0.6–0.8. Additionally, we use weight functions to reconstruct 2D fast-ion distribution functions, given two projected 1D velocity distribution functions from simulated simultaneous measurements with the back- and forward scattering systems.

Continuous Improvement of H-Mode Discharge Performance with Progressively Increasing Lithium Coatings in the National Spherical Torus Experiment

Abstract: Lithium wall coatings have been shown to reduce recycling, improve energy confinement, and suppress edge localized modes in the National Spherical Torus Experiment. Here, we show that these effects depend continuously on the amount of predischarge lithium evaporation. We observed a nearly monotonic reduction in recycling, decrease in electron transport, and modification of the edge profiles and stability with increasing lithium. These correlations challenge basic expectations, given that even the smallest coatings exceeded that needed for a nominal thickness of the order of the implantation range.

Taming the plasma–material interface with the 'snowflake' divertor in NSTX

Abstract: Steady-state handling of divertor heat flux is a critical issue for ITER and future conventional and spherical tokamaks with compact high-power density divertors. A novel 'snowflake' divertor (SFD) configuration was theoretically predicted to have significant magnetic geometry benefits for divertor heat flux mitigation, such as an increased plasma-wetted area and a higher divertor volume available for volumetric power and momentum loss processes, as compared with the standard divertor. Both a significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with core H-mode confinement in discharges with the SFD using only a minimal set of poloidal field coils.

Collisionality and safety factor scalings of H-mode energy transport in the MAST spherical tokamak

Abstract: A factor of 4 dimensionless collisionality scan of H-mode plasmas in MAST shows that the thermal energy confinement time scales as . Local heat transport is dominated by electrons and is consistent with the global scaling. The neutron rate is in good agreement with the ν* dependence of τE,th. The gyrokinetic code GYRO indicates that micro-tearing turbulence might explain such a trend. A factor of 1.4 dimensionless safety factor scan shows that the energy confinement time scales as . These two scalings are consistent with the dependence of energy confinement time on plasma current and magnetic field. Weaker qeng and stronger ν* dependences compared with the IPB98y2 scaling could be favourable for an ST-CTF device, in that it would allow operation at lower plasma current.

Gyrokinetic simulation of momentum transport with residual stress from diamagnetic level velocity shears

Abstract: Residual stress refers to the remaining toroidal angular momentum (TAM) flux (divided by major radius) when the shear in the equilibrium fluid toroidal velocity (and the velocity itself) vanishes. Previously [Waltz et al., Phys. Plasmas 14, 122507 (2007); errata 16, 079902 (2009)], we demonstrated with GYRO [Candy and Waltz, J. Comp. Phys. 186, 545 (2003)] gyrokinetic simulations that TAM pinching from (ion pressure gradient supported or diamagnetic level) equilibrium E×B velocity shear could provide some of the residual stress needed to support spontaneous toroidal rotation against normal diffusive loss. Here we show that diamagnetic level shear in the intrinsic drift wave velocities (or “profile shear” in the ion and electron density and temperature gradients) provides a comparable residual stress. The individual signed contributions of these small (rho-star level) E×B and profile velocity shear rates to the turbulence level and (rho-star squared) ion energy transport stabilization are additive if the r...

Lower Hybrid Current Drive at High Density in Alcator C-Mod

Abstract: Experimental observations of lower hybrid current drive (LHCD) at high density on the Alcator C-Mod tokamak are presented in this paper. Bremsstrahlung emission from relativistic fast electrons in the core plasma drops suddenly above line-averaged densities of 1020 m−3 (ω/ωLH ~ 3) in single null discharges with large (≥8 mm) inner gaps, well below the density limit previously observed on limited tokamaks (ω/ωLH ~ 2). Modelling and experimental evidence suggest that the absence of LHCD driven fast electrons at high density may be due to parasitic collisional absorption in the scrape-off layer (SOL). Experiments show that the population of fast electrons produced by LHCD at high density ( 10^{20}\,{\rm m}^{-3} SRC=http://ej.iop.org/images/0029-5515/51/8/083032/nf381190in001.gif/>) can be increased by operating with an inner gap of less than ~5 mm with the strongest non-thermal emission in inner wall limited plasmas. A change in plasma topology from single to double null produces a modest increase in non-thermal emission at high density. Increasing the electron temperature in the periphery of the plasma (0.8 > r/a > 1.0) also results in a modest increase in non-thermal electron emission above the density limit. Ray tracing/Fokker–Planck simulations of these discharges predict the observed sensitivity to plasma position when the effects of collisional absorption in the SOL are included in the model.

Overview of KSTAR initial operation

Abstract: Since the successful first plasma generation in the middle of 2008, three experimental campaigns were successfully made for the KSTAR device, accompanied with a necessary upgrade in the power supply, heating, wall-conditioning and diagnostic systems. KSTAR was operated with the toroidal magnetic field up to 3.6 T and the circular and shaped plasmas with current up to 700 kA and pulse length of 7 s, have been achieved with limited capacity of PF magnet power supplies.The mission of the KSTAR experimental program is to achieve steady-state operations with high performance plasmas relevant to ITER and future reactors. The first phase (2008–2012) of operation of KSTAR is dedicated to the development of operational capabilities for a super-conducting device with relatively short pulse. Development of start-up scenario for a super-conducting tokamak and the understanding of magnetic field errors on start-up are one of the important issues to be resolved. Some specific operation techniques for a super-conducting device are also developed and tested. The second harmonic pre-ionization with 84 and 110 GHz gyrotrons is an example. Various parameters have been scanned to optimize the pre-ionization. Another example is the ICRF wall conditioning (ICWC), which was routinely applied during the shot to shot interval.The plasma operation window has been extended in terms of plasma beta and stability boundary. The achievement of high confinement mode was made in the last campaign with the first neutral beam injector and good wall conditioning. Plasma control has been applied in shape and position control and now a preliminary kinetic control scheme is being applied including plasma current and density. Advanced control schemes will be developed and tested in future operations including active profiles, heating and current drives and control coil-driven magnetic perturbation.

Instabilities in magnetized spherical Couette flow.

Abstract: We report three-dimensional numerical simulations of the flow of an electrically conducting fluid in a spherical shell when a magnetic field is applied. Different spherical Couette configurations are investigated by varying the rotation ratio between the inner and the outer sphere, the geometry of the imposed field, and the magnetic boundary conditions on the inner sphere. Either a Stewartson layer or a Shercliff layer, accompanied by a radial jet, can be generated depending on the rotation speeds and the magnetic-field strength, and various nonaxisymmetric destabilizations of the flow are observed. We show that instabilities arising from the presence of boundaries present striking similarities with the magnetorotational instability (MRI). To this end, we compare our numerical results to experimental observations of the Maryland experiment [D. R. Sisan et al., Phys. Rev. Lett. 93, 114502 (2004)], which claimed to observe MRI in a similar setup.

ISABELA-QA: query-driven analytics with ISABELA-compressed extreme-scale scientific data

Abstract: Efficient analytics of scientific data from extreme-scale simulations is quickly becoming a top-notch priority. The increasing simulation output data sizes demand for a paradigm shift in how analytics is conducted. In this paper, we argue that query-driven analytics over compressed — rather than original, full-size — data is a promising strategy in order to meet storage-and-I/O-bound application challenges. As a proof-of-principle, we propose a parallel query processing engine, called ISABELA-QA that is designed and optimized for knowledge priors driven analytical processing of spatio-temporal, multivariate scientific data that is initially compressed, in situ, by our ISABELA technology. With ISABELA-QA, the total data storage requirement is less than 23%–30% of the original data, which is upto eight-fold less than what the existing state-of-the-art data management technologies that require storing both the original data and the index could offer. Since ISABELA-QA operates on the metadata generated by our compression technology, its underlying indexing technology for efficient query processing is light-weight; it requires less than 3% of the original data, unlike existing database indexing approaches that require 30%–300% of the original data. Moreover, ISABELA-QA is specifically optimized to retrieve the actual values rather than spatial regions for the variables that satisfy user-specified range queries — a functionality that is critical for high-accuracy data analytics. To the best of our knowledge, this is the first techology that enables query-driven analytics over the compressed spatio-temporal floating-point double-or single-precision data, while offering a light-weight memory and disk storage footprint solution with parallel, scalable, multi-node, multi-core, GPU-based query processing.

Reduced model simulations of the scrape-off-layer heat-flux width and comparison with experiment

Abstract: Reduced model simulations of turbulence in the edge and scrape-off-layer (SOL) region of a spherical torus or tokamak plasma are employed to address the physics of the scrape-off-layer heat-flux width. The simulation model is an electrostatic two-dimensional fluid turbulence model, applied in the plane perpendicular to the magnetic field at the outboard midplane of the torus. The model contains curvature-driven-interchange modes, sheath losses, and both perpendicular turbulent diffusive and convective (blob) transport. These transport processes compete with classical parallel transport to set the SOL width. Midplane SOL profiles of density, temperature, and parallel heat flux are obtained from the simulation and compared with experimental results from the National Spherical Torus Experiment [S. M. Kaye et al., Phys. Plasmas 8, 1977 (2001)] to study the scaling of the heat-flux width with power and plasma current. It is concluded that midplane turbulence is the main contributor to the SOL heat-flux width f...

Effects of Complex Symmetry-Breakings on Alpha Particle Power Loads on First Wall Structures and Equilibrium in ITER

Abstract: Within the ITPA Topical Group on Energetic Particles, we have investigated the impact that various mechanisms breaking the tokamak axisymmetry can have on the fusion alpha particle confinement in ITER as well as on the wall power loads due to these alphas. In addition to the well-known TF ripple, the 3D effect due to ferromagnetic materials (in ferritic inserts and test blanket modules) and ELM mitigation coils are included in these mechanisms. ITER scenario 4 was chosen since, due to its lower plasma current, it is more vulnerable for various off-normal features. First, the validity of using a 2D equilibrium was investigated: a 3D equilibrium was reconstructed using the VMEC code, and it was verified that no 3D equilibrium reconstruction is needed but it is sufficient to add the vacuum field perturbations onto an axisymmetric equilibrium. Then the alpha particle confinement was studied using three independent codes, ASCOT, DELTA5D and F3D OFMC, all of which assume MHD quiescent background plasma and no anomalous diffusion. All the codes gave a loss power fraction of about 0.2%. The distribution of the peak power load was found to depend on the first wall shape. We also made the first attempt to accommodate the effect of fast-ion-related MHD on the wall loads in ITER using the HMGC and ASCOT codes. The power flux to the wall was found to increase due to the redistribution of fast ions by the MHD activity. Furthermore, the effect of the ELM mitigation field on the fast-ion confinement was addressed by simulating NBI ions with the F3D OFMC code. The loss power fraction of NBI ions was found to increase from 0.3% without the ELM mitigation field to 4‐5% with the ELM mitigation field. (Some figures in this article are in colour only in the electronic version)

High spatial sampling global mode structure measurements via multichannel reflectometry in NSTX

Abstract: Global modes?including kinks and tearing modes (f ~ 400?kHz)?play critical roles in many aspects of plasma performance. Their investigation on NSTX is aided by an array of fixed-frequency quadrature reflectometers used to determine their radial density perturbation structure. The array has been recently upgraded to 16 channels spanning 30?75?GHz (ncutoff = (1.1?6.9) ? 1019?m?3 in O-mode), improving spatial sampling and access to the core of H-mode plasmas. The upgrade has yielded significant new results that advance the understanding of global modes in NSTX. The GAE and CAE structures have been measured for the first time in the core of an NSTX high-power (6?MW) beam-heated H-mode plasma. The CAE structure is strongly core-localized, which has important implications for electron thermal transport. The TAE structure has been measured with greatly improved spatial sampling, and measurements of the TAE phase, the first in NSTX, show strong radial variation near the midplane, indicating radial propagation caused by non-ideal MHD effects. Finally, the tearing mode structure measurements provide unambiguous evidence of coupling to an external kink.

Gyrokinetic toroidal simulations on leading multi- and manycore HPC systems

Abstract: The gyrokinetic Particle-in-Cell (PIC) method is a critical computational tool enabling petascale fusion simulation research. In this work, we present novel multi- and manycore-centric optimizations to enhance performance of GTC, a PIC-based production code for studying plasma microturbulence in tokamak devices. Our optimizations encompass all six GTC sub-routines and include multi-level particle and grid decompositions designed to improve multi-node parallel scaling, particle binning for improved load balance, GPU acceleration of key subroutines, and memory-centric optimizations to improve single-node scaling and reduce memory utilization. The new hybrid MPI-OpenMP and MPI-OpenMP-CUDA GTC versions achieve up to a 2x speedup over the production Fortran code on four parallel systems --- clusters based on the AMD Magny-Cours, Intel Nehalem-EP, IBM BlueGene/P, and NVIDIA Fermi architectures. Finally, strong scaling experiments provide insight into parallel scalability, memory utilization, and programmability trade-offs for large-scale gyrokinetic PIC simulations, while attaining a 1.6× speedup on 49,152 XE6 cores.

Characterization of intrinsic rotation drive on DIII-D

Abstract: Recent experiments on DIII-D have focused on elucidating the drive mechanisms for intrinsic rotation in tokamak fusion plasmas. In H-mode plasmas, the effective torque at the edge (ρ > 0.8) associated with the intrinsic rotation shows a dependence on the pedestal pressure gradient ∇Pped, which is qualitatively consistent with models describing E × B shear as a means of creating 'residual stress' and driving intrinsic rotation. However, direct measurement of the turbulent Reynolds stress using probes suggests that this is not the full picture. Specifically, there is a significant mismatch between the plasma spin up and the inferred torque from the Reynolds stress at the edge, indicating that additional mechanisms are necessary to completely understand edge intrinsic rotation generation. A narrow rotation layer is observed near the separatrix, which can qualitatively be explained using a model of thermal ion orbit loss. Parametrically, the torque from such a process is expected to vary with . A good predictor of the edge intrinsic torque is obtained by including this dependence, together with the previously observed ∇Pped dependence, in a regression fit of a wide range of H-mode conditions. The intrinsic torque in the core (ρ < 0.5) of H-mode plasmas tends to be much smaller than observed at the edge, although some examples have been found where it is large enough to modify the rotation profile. For instance, in certain plasmas with electron cyclotron heating, a significant counter-intrinsic torque has been observed in the inner region of the plasma.

Requirements on localized current drive for the suppression of neoclassical tearing modes

Abstract: A heuristic criterion for the full suppression of an NTM was formulated as ηNTM ≡ jCD,max/jBS ≥ 1.2 (Zohm et al 2005 J. Phys. Conf. Ser. 25 234), where jCD,max is the maximum in the driven current density profile applied to stabilize the mode and jBS is the local bootstrap current density. In this work we subject this criterion to a systematic theoretical analysis on the basis of the generalized Rutherford equation. Taking into account only the effect of jCD inside the island, a new criterion for full suppression by a minimum applied total current is obtained in the form of a maximum allowed value for the width of the driven current, wdep, combined with a required minimum for the total driven current in the form of wdepηNTM, where both limits depend on the marginal and saturated island sizes. These requirements can be relaxed when additional effects are taken into account, such as a change in the stability parameter Δ' from the current driven outside the island, power modulation, the accompanying heating inside the island or when the current drive is applied preemptively. When applied to ITER scenario 2, the requirement for full suppression of either the 3/2 or 2/1 NTM becomes wdep 5 cm and wdepηNTM 5 cm in agreement with (Sauter et al 2010 Plasma Phys. Control. Fusion 52 025002). Optimization of the ITER ECRH Upper Port Launcher design towards minimum required power for full NTM suppression requires an increase in the toroidal injection angle of the lower steering mirror of several degrees compared with its present design value, while for the upper steering mirror the present design value is close to the optimum.

Integrated modelling of steady-state scenarios and heating and current drive mixes for ITER

Abstract: Recent progress on ITER steady-state (SS) scenario modelling by the ITPA-IOS group is reviewed. Code-to-code benchmarks as the IOS group's common activities for the two SS scenarios (weak shear scenario and internal transport barrier scenario) are discussed in terms of transport, kinetic profiles, and heating and current drive (CD) sources using various transport codes. Weak magnetic shear scenarios integrate the plasma core and edge by combining a theory-based transport model (GLF23) with scaled experimental boundary profiles. The edge profiles (at normalized radius rho = 0.8-1.0) are adopted from an edge-localized mode-averaged analysis of a DIII-D ITER demonstration discharge. A fully noninductive SS scenario is achieved with fusion gain Q = 4.3, noninductive fraction f(NI) = 100%, bootstrap current fraction f(BS) = 63% and normalized beta beta(N) = 2.7 at plasma current I(p) = 8MA and toroidal field B(T) = 5.3 T using ITER day-1 heating and CD capability. Substantial uncertainties come from outside the radius of setting the boundary conditions (rho = 0.8). The present simulation assumed that beta(N)(rho) at the top of the pedestal (rho = 0.91) is about 25% above the peeling-ballooning threshold. ITER will have a challenge to achieve the boundary, considering different operating conditions (T(e)/T(i) approximatemore » to 1 and density peaking). Overall, the experimentally scaled edge is an optimistic side of the prediction. A number of SS scenarios with different heating and CD mixes in a wide range of conditions were explored by exploiting the weak-shear steady-state solution procedure with the GLF23 transport model and the scaled experimental edge. The results are also presented in the operation space for DT neutron power versus stationary burn pulse duration with assumed poloidal flux availability at the beginning of stationary burn, indicating that the long pulse operation goal (3000s) at I(p) = 9 MA is possible. Source calculations in these simulations have been revised for electron cyclotron current drive including parallel momentum conservation effects and for neutral beam current drive with finite orbit and magnetic pitch effects.« less

Density gradient stabilization of electron temperature gradient driven turbulence in a spherical tokamak.

Abstract: In this letter we report the first clear experimental observation of density gradient stabilization of electron temperature gradient driven turbulence in a fusion plasma. It is observed that longer wavelength modes, k⊥ρs ≤10, are most stabilized by density gradient, and the stabilization is accompanied by about a factor of two decrease in the plasma effective thermal diffusivity.

The relationships between edge localized modes suppression, pedestal profiles and lithium wall coatings in NSTX

Abstract: Recently in the National Spherical Torus Experiment (NSTX), increasing lithium wall coatings suppressed edge localized modes (ELMs), gradually but not quite monotonically. This work details profile and stability analysis as ELMs disappeared throughout the lithium scan. While the quantity of lithium deposited between discharges did not uniquely determine the presence of ELMs, profile analysis demonstrated that lithium was correlated with wider density and pressure pedestals with peak gradients farther from the separatrix. Moreover, the ELMy and ELM-free discharges were cleanly separated by their density and pedestal widths and peak gradient locations. Ultimately, ELMs were only suppressed when lithium caused the density pedestal to widen and shift inward. These changes in the density gradient were directly reflected in the pressure gradient and calculated bootstrap current. This supports the theory that ELMs in NSTX are caused by peeling and/or ballooning modes, as kink/peeling modes are stabilized when the edge current and pressure gradient shift away from the separatrix. Edge stability analysis using ELITE corroborated this picture, as reconstructed equilibria from ELM-free discharges were generally farther from their kink/peeling stability boundaries than ELMy discharges. We conclude that density profile control provided by lithium is the key first step to ELM suppression in NSTX.

Evidence for the importance of trapped particle resonances for resistive wall mode stability in high beta tokamak plasmas.

Abstract: Active measurements of the plasma stability in tokamak plasmas reveal the importance of kinetic resonances for resistive wall mode stability. The rotation dependence of the magnetic plasma response to externally applied quasistatic n=1 magnetic fields clearly shows the signatures of an interaction between the resistive wall mode and the precession and bounce motions of trapped thermal ions, as predicted by a perturbative model of plasma stability including kinetic effects. The identification of the stabilization mechanism is an essential step towards quantitative predictions for the prospects of "passive" resistive wall mode stabilization, i.e., without the use of an "active" feedback system, in fusion-alpha heated plasmas. © 2011 American Physical Society.

Overview of the physics and engineering design of NSTX upgrade

Abstract: The spherical tokamak (ST) is a leading candidate for a fusion nuclear science facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the U.S. actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3–6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1–1.5s to 5s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high β N , and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up (from I P = 0.4MA up to 0.8–1MA) as needed for an ST-based FNSF. In boundary physics, NSTX and higher-A tokamaks measure an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavorably impact next-step devices. Recently, NSTX has successfully demonstrated very high flux expansion and substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade are described.