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At least 19 records

Fast wave heating in the NSTX-Upgrade device

NSTX-Upgrade will operate with toroidal magnetic fields (B{sub T}) up to 1 T, nearly twice the value used in the experiments on NSTX, and the available NBI power will be doubled. The doubling of B{sub T} while retaining the 30 MHz RF source frequency has moved the heating regime from the high harmonic fast wave (HHFW) regime used in NSTX to the mid harmonic fast wave (MHFW) regime. By making use of the full wave code AORSA, this work shows that direct ion damping (mainly by thermal ions localized at the 5th harmonic resonance) might be significant in NSTX-Upgrade under TRANSP predicted full performance conditions and the electron and ion absorption is sensitive to the ratio of electron and ion temperature. Launching at high toroidal wave number appears to be one way to significantly reduce the ion damping. By using the extended AORSA code, which includes a detailed description of the scrape-off layer in the field solutions, we found a large electric field amplitude outside of the last closed flux surface as previously seen in NSTX from AORSA simulations (D. L. Green, et al, Phys. Rev. Lett. 107, 145001 (2011)). Preliminary results by introducing a collision damping in the scrape-off layer in the AORSA code to represent a damping process are presented, showing for the first time absorbed power in the scrape-off layer.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Propagation And Damping Of High Harmonic Fast Waves And Electron Cyclotron Waves In The Nstx-U-Device

For the spherical torus and for conventional tokamaks to be viable means of producing fusion energy there are two important requirements. The first is to heat the plasma in the experimental devices to high temperatures; preferably, to fusion relevant temperatures. The second is to drive a current in the plasma so that the fusion device operates in a steady-state mode. A pulsed fusion reactor for energy production is not desirable. A favored means to achieve both of these requirements is through the use of microwaves, commonly referred to as radio frequency waves. This proposal was to understand observations from the National Spherical Torus Experiment Upgrade (NSTX-U) device on radio frequency wave heating of the confined plasma during operations with high harmonic fast waves. The proposed research was also directed toward assisting in the optimization of the experimental modes of operation so as to maximize the heating efficiency. A detailed analysis of the experiments has shown that about one-third of the input radio frequency power does not make it into the core region of the plasma where heating is required. The losses are believed to occur primarily in the edge region where the radio frequency power is initially coupled into the plasma. Using sophisticated computational codes and theoretical models, that have been developed for the NSTX-U experimental program, this research quantified the physical processes which occur in the edge plasma and lead to a decrease in the heating efficiency. The effect of edge turbulence on the scattering of radio frequency waves was studied for waves in the ion cyclotron and electron cyclotron range of frequencies. For high harmonic fast waves in the ion cyclotron range of frequencies, simulations were carried out using a full-wave solver that was extended to the vacuum vessel wall. These simulations successfully reproduced trends that were observed experimentally. This simulation capability was also used to study the interaction of high harmonic fast waves with the scrape-off layer plasma. More recently, a new technique has been developed for the calculation of radio frequency waves in toroidal geometry that enables the simultaneous incorporation of antenna geometry, plasma facing components, the scrape off-layer, and core propagation in the high harmonic fast wave regime. The effect of turbulent plasmas on the mode conversion of extraordinary electromagnetic waves in the electron cyclotron range of frequencies in the NSTX-U plasma has been studied theoretically. In the over-dense plasmas of NSTX-U, the traditional electromagnetic waves are not effective for delivering energy and momentum to the core plasma. However, electron Bernstein waves generated by mode conversion are efficient carriers of wave energy and momentum to the core. The mode conversion occurs in the turbulent edge region. A theory for mode conversion in such plasmas has been developed. The parameters that can optimize the mode conversion can be readily determined. Also, the model is useful for experimentalists to analyze observation on a shot-to-shot basis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Propagation And Damping Of High Harmonic Fast Waves And Electron Cyclotron Waves In The Nstx-U-Device

For the spherical torus and for conventional tokamaks to be viable means of producing fusion energy there are two important requirements. The first is to heat the plasma in the experimental devices to high temperatures; preferably, to fusion relevant temperatures. The second is to drive a current in the plasma so that the fusion device operates in a steady-state mode. A pulsed fusion reactor for energy production is not desirable. A favored means to achieve both of these requirements is through the use of microwaves, commonly referred to as radio frequency waves. This proposal was to understand observations from the National Spherical Torus Experiment Upgrade (NSTX-U) device on radio frequency wave heating of the confined plasma during operations with high harmonic fast waves. The proposed research was also directed toward assisting in the optimization of the experimental modes of operation so as to maximize the heating efficiency. A detailed analysis of the experiments has shown that about one-third of the input radio frequency power does not make it into the core region of the plasma where heating is required. The losses are believed to occur primarily in the edge region where the radio frequency power is initially coupled into the plasma. Using sophisticated computational codes and theoretical models, that have been developed for the NSTX-U experimental program, this research quantified the physical processes which occur in the edge plasma and lead to a decrease in the heating efficiency. The effect of edge turbulence on the scattering of radio frequency waves was studied for waves in the ion cyclotron and electron cyclotron range of frequencies. For high harmonic fast waves in the ion cyclotron range of frequencies, simulations were carried out using a full-wave solver that was extended to the vacuum vessel wall. These simulations successfully reproduced trends that were observed experimentally. This simulation capability was also used to study the interaction of high harmonic fast waves with the scrape-off layer plasma. More recently, a new technique has been developed for the calculation of radio frequency waves in toroidal geometry that enables the simultaneous incorporation of antenna geometry, plasma facing components, the scrape off-layer, and core propagation in the high harmonic fast wave regime. The effect of turbulent plasmas on the mode conversion of extraordinary electromagnetic waves in the electron cyclotron range of frequencies in the NSTX-U plasma has been studied theoretically. In the over-dense plasmas of NSTX-U, the traditional electromagnetic waves are not effective for delivering energy and momentum to the core plasma. However, electron Bernstein waves generated by mode conversion are efficient carriers of wave energy and momentum to the core. The mode conversion occurs in the turbulent edge region. A theory for mode conversion in such plasmas has been developed. The parameters that can optimize the mode conversion can be readily determined. Also, the model is useful for experimentalists to analyze observation on a shot-to-shot basis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

The Center for Momentum Transport and Flow Organization in Plasmas - Final Scientific Report

Overview of University of Colorado Efforts: The University of Colorado group has focused on two primary fronts during the grant period: development of a variety of multi-point diagnostic and/or imaging analysis techniques, and momentum-transport related experiments on a variety of devices (NSTX at PPPL, CSDX at UCSD, LAPD at UCLA, DIII-D at GA). Experimental work has taken advantage of several diagnostic instruments, including fast-framing cameras for imaging of electron density fluctuations (either directly or using injected gas puffs), ECEI for imaging of electron temperature fluctuations, and multi-tipped Langmuir and magnetic probes for corroborating measurements of Reynolds and Maxwell stresses. Mode Characterization in CSDX: We have performed a series of experiments at the CSDX linear device at UCSD, in collaboration with Center PI G. Tynan's group. The experiments included a detailed study of velocity estimation techniques, including direct comparisons between Langmuir probes and image-based velocimetry from fast-framing camera data. We used the camera data in a second set of studies to identify the spatial and spectral structure of coherent modes, which illuminates wave behavior to a level of detail previously unavailable, and enables direct comparison of dispersion curves to theoretical estimates. In another CSDX study, similar techniques were used to demonstrate a controlled transition from nonlinearly coupled discrete eigenmodes to fully developed broadband turbulence. The axial magnetic field was varied from 40-240 mT, which drove the transition. At low magnetic fields, the plasma is dominated by drift waves. As the magnetic field is increased, a strong potential gradient at the edge introduces an ExB shear-driven instability. At the transition, another mode with signatures of a rotation-induced Rayleigh–Taylor instability appears at the central plasma region. Concurrently, large axial velocities were found in the plasma core. For larger magnetic fields, all of the instabilities co-exist, leading to rich plasma dynamics and fully developed broadband turbulence. Edge-Turbulence and Flow Experiments in NSTX: A series of Gas Puff Imaging (GPI) observations on NSTX revealed a quasi-periodic oscillation in the plasma edge preceding the L-H transition in a limited set of neutral beam heated plasmas. These ~3 kHz flow oscillations exhibit both long wavelength and long correlation lengths, suggesting they are zonal-flow-like. The flow oscillations are strongly correlated with modulations of the level of edge turbulence, thus the system appears to undergo a predator--prey-type limit-cycle preceding the L-H transition. However, a clear trigger for the L-H transition was not observed. Reynolds stress profiles were obtained directly from image velocimetry for L-mode periods ELM-Precursor Studies in NSTX: A separate study based on NSTX-GPI data captured the two-dimensional evolution of edge-localized mode (ELM) precursors. Precursor events were observed preceding ELMs and ELM-induced H–L back-transitions in radio-frequency heated H-mode plasmas, and the growth of the precursor mode through the ELM filamentation was imaged in the plane perpendicular to the local B-field. Strong edge intensity modulations appeared to propagate in the electron diamagnetic direction while steadily drifting radially outwards. Intensity fluctuations were observed at frequencies around 20 kHz and wavenumbers of 0.05-0.2 cm -1 . Upon growing to a trigger point, precursor fluctuations were seen to form filamentary structures and move into the scrape-off layer (SOL) explosively with radial velocities peaking at 8 km/s. Once in the SOL, filaments reverse their propagation direction and travel in the ion diamagnetic direction. Edge intensity fluctuations were strongly correlated with magnetic signals from Mirnov coils, and toroidally distributed coils estimated toroidal mode numbers of n=5-10. Quantitatively similar precursors have been observed in ohmic H-mode plasmas as well, though significantly fewer events are seen in the ohmic cases and none were observed in the near-threshold NBI H-modes studied. Development of Velocimetry and Image Analysis Techniques: Along with the experiments listed above, the Colorado group has continued to explore various velocimetry techniques and their range of validity. We have developed a “linear optical flow” code, which calculates smooth velocity maps while accurately assessing local regions of high curl. This is critical for separating spatial scales of velocity behavior, and thus transport. This code has recently come on-line, and we are currently using it to revisit a number of older datasets. Additionally, we have worked on developing pattern-recognition techniques for imaging diagnostics, based on established digital image compression algorithms. This has the potential to open the analysis of turbulent plasma behavior beyond the well-trodden Fourier and wavelet approaches. Finally, we have extended several of these image-analysis routines to multiple other diagnostic sets, including GPI datasets in NSTX and ECEI imaging on DIII-D. In one study, correlation maps were used to measure the full 2-D mode structure of drift-wave level structures using ECEI for the first time. This enabled direct comparison to gyrokinetic simulations using the GEM code, from which is was determined that TEM modes were being measured in DIII-D.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Bus Design for the Poloidal Field Coils of the NSTX-Upgrade Fusion Device

The National Spherical Torus eXperiment (NSTX) has undergone a major upgrade to NSTX-U at Princeton Plasma Physics Laboratory (PPPL). NSTX upgrade (NSTX-U) will double the toroidal field, plasma current, and neutral beam injection heating power, as well as significantly increase the pulse duration. NSTX-U uses three poloidal field (PF) coils at the vessel top and three at the bottom near the divertor areas to control the local plasma shape there. These shaping coils operate at a maximum terminal voltage of 2 kV, corresponding to a maximum current about 20 kA. To supply the power to the PF coils, hard copper bus bars are typically used near the vacuum vessel, while water cooled flexible cables are used away from the vessel. The detailed design of the hard copper bus bars will be covered in this article. During operation, these hard bus bars are subject to high electromagnetic (EM) forces, thermal displacement loads, as well as plasma halo loads due to disruption. The EM, thermal, and structural analysis were performed, and the results revealed that, with the worst loads combined, the new design meets the NSTX-U thermal, structural, and fatigue cycle requirements. Finally, the manufacturing and installation process of the bus work will be discussed at the end of this article.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Edge and Scrape-off-Layer Physics for NSTX-U (Final Report)

U.S. Department of Energy (DOE) supported work at Lodestar Research Corporation under grant DE-FG02-02ER54678 is summarized in this final report. The research conducted under this grant addressed key scientific issues associated with the plasma confinement and fusion energy experimental devices NSTX and its upgrade NSTX-U located at Princeton Plasma Physics Laboratory. Harnessing the energy that is released in fusion reactions would provide a safe and abundant source of power to meet the growing energy needs of the world population. Lodestar investigated the important cm-scale region near the edge of these devices where hot plasma, close to one million degrees, meets material surfaces in what is known as a scrape-off layer (SOL). This critical region is subject to plasma turbulence and plasma flows which must be understood to optimize fusion performance on future machines. The work employed theoretical and theory-based computational models for validation and applied the models to understand experimental results.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Design Considerations for the Implementation of a High-Field-Side Transient CHI System on QUEST

Transient coaxial helicity injection (T-CHI), a method first developed on the small helicity injected torus (HIT-II) experiment and then validated on the much larger National Spherical Torus Experiment (NSTX) device, is a method to initiate an inductive-like tokamak plasma discharge without reliance on the central solenoid. A CHI discharge is initiated by driving current along magnetic flux that connects the inner and outer divertor plates on one end of the tokamak. To permit this, on both HIT-II and NSTX, toroidal ceramic insulators were used to electrically separate the inner and outer vessel components. The use of such large toroidal vacuum insulators may not be easy to implement in reactors. To address this issue, the Q-shu University experiments with steady-state spherical tokamak (ST) (QUEST) is developing a reactor-relevant CHI configuration in which one of the divertor plates is electrically insulated from the rest of the vessel. The first application of T-CHI on QUEST biased the CHI electrode to the outer vessel. While the CHI discharges could be easily generated, it was found that as the discharge filled the vessel, the separation distance between the injector magnetic flux footprints widened, a condition that is not favorable for the generation of closed flux surfaces. Biasing the electrode to the inner wall is a configuration similar to that used on NSTX and HIT-II, but initial testing in this configuration has proved to be challenging. The design described here overcomes the present limitation by locating the CHI electrode much closer to the CHI injector flux coil and using an NSTX-like gas injection manifold to enable high-field-side T-CHI startup on QUEST. Finally, the concepts described in this article should also benefit the future implementation of T-CHI systems in other tokamaks and spherical tokamaks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

The Role of Lithium Conditioning in Achieving High Performance, Long Pulse H-mode Discharges in the NSTX and EAST Devices

In this paper, the role of lithium wall conditioning on the achievement of high performance, long pulse discharges in the National Spherical Torus Experiment (NSTX) and the Experimental Advanced Superconducting Tokamak (EAST) is documented. Common observations include recycling reduction and elimination of ELMs. In NSTX, lithium conditioning typically resulted in ELM-free operation with impurity accumulation, which was ameliorated e.g. with pulsed 3D fields to trigger controlled ELMs. Active lithium conditioning in EAST discharges has overcome this problem, producing an ELM-free Hmode with controlled density and impurities.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Properties of Alfvén eigenmodes in the Toroidal Alfvén Eigenmode range on the National Spherical Torus Experiment-Upgrade

A second Neutral Beam (NB) injection line is being installed on the NSTX Upgrade device, resulting in six NB sources with different tangency radii that will be available for heating and current drive. This work explores the properties of instabilities in the frequency range of the Toroidal Alfvén Eigenmode (TAE) for NSTX-U scenarios with various NB injection geometries, from more perpendicular to more tangential, and with increased toroidal magnetic field with respect to previous NSTX scenarios. Predictions are based on analysis through the ideal MHD code NOVA-K. For the scenarios considered in this work, modifications of the Alfvén continuum result in a frequency up-shift and a broadening of the radial mode structure. The latter effect may have consequences for fast ion transport and loss. Preliminary stability considerations indicate that TAEs are potentially unstable with ion Landau damping representing the dominant damping mechanism.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Conceptual design of a divertor Thomson scattering diagnostic for NSTX-U

A conceptual design for a divertor Thomson scattering (DTS) diagnostic has been developed for the NSTX-U device to operate in parallel with the existing multipoint Thomson scattering system. Higher projected peak heat flux in NSTX-U will necessitate application of advanced magnetics geometries and divertor detachment. Interpretation and modeling of these divertor scenarios will depend heavily on local measurement of electron temperature, T{sub e}, and density, n{sub e}, which DTS provides in a passive manner. The DTS design for NSTX-U adopts major elements from the successful DIII-D DTS system including 7-channel polychromators measuring T{sub e} to 0.5 eV. If implemented on NSTX-U, the divertor TS system would provide an invaluable diagnostic for the boundary program to characterize the edge plasma.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

TRANSP-based closed-loop simulations of current profile optimal regulation in NSTX-Upgrade

Active control of the toroidal current density profile is critical for the upgraded National Spherical Torus eXperiment device (NSTX-U) to maintain operation at the desired high-performance, MHD-stable, plasma regime. Initial efforts towards current density profile control have led to the development of a control-oriented, physics-based, plasma-response model, which combines the magnetic diffusion equation with empirical correlations for the kinetic profiles and the non-inductive current sources. The developed control-oriented model has been successfully tailored to the NSTX-U geometry and actuators. Moreover, a series of efforts have been made towards the design of model-based controllers, including a linear-quadratic-integral optimal control strategy that can regulate the current density profile around a prescribed target profile while rejecting disturbances. In this work, the tracking performance of the proposed current-profile optimal controller is tested in numerical simulations based on the physics-oriented code TRANSP. Furthermore, these high-fidelity closed-loop simulations, which are a critical step before experimental implementation and testing, are enabled by a flexible framework recently developed to perform feedback control design and simulation in TRANSP.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

TRANSP-based closed-loop simulations of current profile optimal regulation in NSTX-Upgrade

Active control of the toroidal current density profile is critical for the upgraded National Spherical Torus eXperiment device (NSTX-U) to maintain operation at the desired high-performance, MHD-stable, plasma regime. Initial efforts towards current density profile control have led to the development of a control-oriented, physics-based, plasma-response model, which combines the magnetic diffusion equation with empirical correlations for the kinetic profiles and the non-inductive current sources. The developed control-oriented model has been successfully tailored to the NSTX-U geometry and actuators. Moreover, a series of efforts have been made towards the design of model-based controllers, including a linear-quadratic-integral optimal control strategy that can regulate the current density profile around a prescribed target profile while rejecting disturbances. In this work, the tracking performance of the proposed current-profile optimal controller is tested in numerical simulations based on the physics-oriented code TRANSP. These high-fidelity closed-loop simulations, which are a critical step before experimental implementation and testing, are enabled by a flexible framework recentlydeveloped to perform feedback control design and simulation in TRANSP.

current profile control↗

NSTX-U theory, modeling and analysis results

Here, the mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 s, and has available up to 15 MW of neutral beam injection power at different tangency radii and 6 MW of high harmonic fast wave heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle (EP) parameter space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, EP stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

A poloidal high- k scattering system for NSTX-U

A previous 5-channel tangential high-k scattering system is being replaced by an 8-channel, poloidal high-k scattering system on the National Spherical Torus eXperiment Upgrade (NSTX-U) device located in Princeton, NJ, USA. The 693 GHz poloidal scattering system replaces a 280 GHz tangential scattering system to study high-k electron density fluctuations on NSTX-U, thereby considerably enhancing planned turbulence physics studies by providing a measurement of the k θ -spectrum of both electron temperature gradient (ETG) and ion temperature gradient (ITG) modes. Two approaches to generating the 693 GHz probe beam are under development: an optically-pumped far-infrared (FIR) laser that generates ~50 mW, and a compact gyrotron that can potentially generate in excess of 5 W. Large aperture optics collect radiation scattered from density fluctuations in the plasma core at 8 simultaneous scattering angles ranging from 2 to 15° corresponding to poloidal wavenumbers that extend to >40 cm -1 . Finally, steerable launch optics coupled with receiver optics mounted on a 5-axis receiver carriage allow the scattering volume to be placed radially from r/a = 0.3 out to the pedestal region (r/a ~ 0.99) and translated horizontally as needed to satisfy wavenumber matching.

47 OTHER INSTRUMENTATION↗

Final Report: Spectral Analysis of L-shell Data in the Extreme Ultraviolet from Tokamak Plasmas

We performed detailed analyses of extreme ultraviolet spectra taken by Lawrence Livermore National Laboratory on the National Spherical Torus Experiment at Princeton Plasma Physics Laboratory and on the Alcator CKmod tokamak at the M.I.T. Plasma Science and Fusion Center. We focused on the emission of iron, carbon, and other elements in several spectral band pass regions covered by the Atmospheric Imaging Assembly on the Solar Dynamics Observatory. We documented emission lines of carbon not found in currently used solar databases and demonstrated that this emission was due to charge exchange.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Properties of Alfvén Eigenmodes in the TAE range on the National Spherical Torus Experiment-Upgrade

A second Neutral Beam (NB) injection line is being installed on the NSTX Upgrade device, resulting in six NB sources with different tangency radii that will be available for heating and current drive. This work explores the properties of instabilities in the frequency range of the Toroidal Alfvén Eigenmode (TAE) for NSTX-U scenarios with various NB injection geometries, from more perpendicular to more tangential, and with increased toroidal magnetic field with respect to previous NSTX scenarios. Predictions are based on analysis through the ideal MHD code NOVA-K. For the scenarios considered in this work, modifications of the Alfvén continuum result in a frequency up-shift and a broadening of the radial mode structure. The latter effect may have consequences for fast ion transport and loss. Preliminary stability considerations indicate that TAEs are potentially unstable, with ion Landau damping representing the dominant damping mechanism.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Energetic particles in spherical tokamak plasmas

Spherical tokamaks (STs) typically have lower magnetic fields than conventional tokamaks, but similar mass densities. Suprathermal ions with relatively modest energies, in particular beam-injected ions, consequently have speeds close to or exceeding the Alfvén velocity, and can therefore excite a range of Alfvénic instabilities which could be driven by (and affect the behaviour of) fusion α-particles in a burning plasma. STs heated with neutral beams, including the small tight aspect ratio tokamak (START), the mega amp spherical tokamak (MAST), the national spherical torus experiment (NSTX) and Globus-M, have thus provided an opportunity to study toroidal Alfvén eigenmodes (TAEs), together with higher frequency global Alfvén eigenmodes (GAEs) and compressional Alfvén eigenmodes (CAEs), which could affect beam current drive and channel fast ion energy into bulk ions in future devices. In NSTX GAEs were correlated with a degradation of core electron energy confinement. In MAST pulses with reduced magnetic field, CAEs were excited across a wide range of frequencies, extending to the ion cyclotron range, but were suppressed when hydrogen was introduced to the deuterium plasma, apparently due to mode conversion at ion–ion hybrid resonances. At lower frequencies fishbone instabilities caused fast particle redistribution in some MAST and NSTX pulses, but this could be avoided by moving the neutral beam line away from the magnetic axis or by operating the plasma at either high density or elevated safety factor. Fast ion redistribution has been observed during GAE avalanches on NSTX, while in both NSTX and MAST fast ions were transported by saturated kink modes, sawtooth crashes, resonant magnetic perturbations and TAEs. The energy dependence of fast ion redistribution due to both sawteeth and TAEs has been studied in Globus-M. High energy charged fusion products are unconfined in present-day STs, but have been shown in MAST to provide a useful diagnostic of beam ion behaviour, supplementing the information provided by neutron detectors. In MAST electrons were accelerated to highly suprathermal energies as a result of edge localised modes, while in both MAST and NSTX ions were accelerated due to internal reconnection events. Lastly, ion acceleration has also been observed during merging-compression start-up in MAST.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Study of plasma equilibrium in toroidal fusion devices using mesh-free numerical calculation method

Plasma confinement using external magnetic field is one of the successful ways leading to the controlled nuclear fusion. Development and validation of the solution process for plasma equilibrium in the experimental toroidal fusion devices is the main subject of this work. Solution of the nonlinear 2D stationary problem as posed by the Grad-Shafranov equation gives quantitative information about plasma equilibrium inside the vacuum chamber of hot fusion devices. This study suggests solving plasma equilibrium equation which is essential in toroidal nuclear fusion devices, using a mesh-free method in a condition that the plasma boundary is unknown. The Grad-Shafranov equation has been solved numerically by the point interpolation collocation mesh-free method. Important features of this approach include truly mesh free, simple mathematical relationships between points and acceptable precision in comparison with the parametric results. The calculation process has been done by using the regular and irregular nodal distribution and support domains with different points. The relative error between numerical and analytical solution is discussed for several test examples such as small size Damavand tokamak, ITER-like equilibrium, NSTX-like equilibrium, and typical Spheromak.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗