Engineering topics

Evans, T E

Publications and source records attributed to Evans, T E.

15 records

DOE OSTI2019

Study of the impact of pre- and real-time deposition of lithium on plasma performance on NSTX

The efficiency of two lithium (Li) injection methods used on the National Spherical Torus Experiment (NSTX) are compared in terms of the amount of Li used to produce equivalent plasma performance improvements, namely Li evaporation over the divertor plates, prior to the initiation of the discharge, and real-time Li injection directly into the plasma scrape-off layer during the discharge. The measurements show that the real-time method can affect the energy confinement and edge stability of NSTX plasmas in a more efficient way than the Li evaporation method as it requires only a fraction of the amount of Li used by the evaporation method to produce similar improvements.

Lithium↗

DOE OSTI2012

Intrinsic rotation produced by ion orbit loss and X-loss

A practical calculation model for the intrinsic rotation imparted to the edge plasma by the directionally preferential loss of ions on orbits that cross the last closed flux surface is presented and applied to calculate intrinsic rotation in several DIII-D [J. Luxon, Nucl. Fusion 42, 614 (2002)] discharges. The intrinsic rotation produced by ion loss is found to be sensitive to the edge temperature and radial electric field profiles, which has implications for driving intrinsic rotation in future large tokamaks.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2012

Impact of screening of resonant magnetic perturbations in three dimensional edge plasma transport simulations for DIII-D

The impact of resonant magnetic perturbations (RMPs) on the plasma edge can be analyzed in detail by three dimensional computer simulations, which take the underlying magnetic field structure as input. Previously, the 'vacuum approximation' has been used to calculate the magnetic field structure although plasma response effects may result in a screening (or even an amplification) of the external perturbations. Simulation results for an ITER similar shape plasma at the DIII-D tokamak are presented for the full vacuum perturbation field and an ad hoc screening case in comparison to the unperturbed configuration. It is shown that the RMP induced helical patterns in the plasma edge and on the divertor target shrink once screening is taken into account. However, a flat temperature profile is still found in the 'open field line domain' inside the separatrix, while the 'density pump out effect' found in the vacuum RMP case is considerably weakened.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2012

Measurements of hard x-ray emission from runaway electrons in DIII-D

The spatial distribution of runaway electron (RE) strikes to the wall during argon pellet-initiated rapid shutdown of diverted and limited plasma shapes in DIII-D is studied using a new array of hard x-ray (HXR) scintillators. Two plasma configurations were investigated: an elongated diverted H-mode and a low-elongation limited L-mode. HXR emission from MeV level REs generated during the argon pellet injection is observed during the thermal quench (TQ) in diverted discharges from REs lost into the divertor. In limiter discharges, this prompt TQ loss is reduced, suggesting improved TQ confinement of REs in this configuration. During the plateau phase when the plasma current is carried by REs, toroidally symmetric HXR emission from remaining confined REs is seen. Transient HXR bursts during this RE current plateau suggest the presence of a small level of wall losses due to the presence of an unidentified instability. Eventually, an abrupt final loss of the remaining RE current occurs. This final loss HXR emission shows a strong toroidal peaking and a consistent spatiotemporal evolution that suggests the development of a kink instability.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2012

Resonant features of energy and particle transport during application of resonant magnetic perturbation fields at TEXTOR and DIII-D

In this paper, results of a direct comparison of TEXTOR and DIII-D experiments with resonant magnetic perturbation (RMP) fields are presented. This comparison of resistive L-mode plasmas at TEXTOR with highly conductive H-mode plasmas at DIII-D is useful to identify generic physics mechanisms during application of RMP fields with a strong field line pitch angle alignment in the plasma edge. A reduction in the pedestal electron pressure p(e) with increasing extension of the vacuum modelled stochastic layer and p(e) recovery with decreasing layer width is found caused by a q(95) resonant reduction in the edge (0.8 < Psi(N) < 0.95) electron temperature T-e(q(95)) on both devices. For RMP edge-localized mode (ELM) suppressed H-mode plasmas at DIII-D, the gradients del T-e and nominal values of T-e are reduced in this edge region while increasing in the pedestal (0.95 < Psi(N) < 1.0) with RMP field applied and both are highly dependent on q(95). In contrast, an increase in the central ion temperature with strong steepening of the ion temperature profile at mid-radius is found-again being highly dependent on q(95). However, these resonant thermal transport effects are only seen in high triangularity plasmas revealing a strong shape dependence of the thermal transport. In contrast to the highly q(95) dependent thermal transport features, the reduction of n(e)-known as density pump out-shows a much weaker dependence on q(95). We show the potential to reduce the RMP induced particle pump out by fine tuning of the RMP spectral properties. At low resonant field amplitudes enhanced particle confinement is seen in high-field side limited L-mode discharges on both devices while higher resonant field amplitudes yield particle pumps out.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2011

Effect of thermoelectric current splitting on the magnetic topology in DIII-D

An extended model is proposed to describe the magnetic topology during appearance of edge localized modes (ELMs). It is applied to an ELMing H-mode in a lower single null discharge at DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)]. The process of flux tube formation is discussed based on a previously proposed two-step model. Large currents, as found in measurements in DIII-D, are assumed running through newly formed large flux tubes. Two different realizations of the current distribution within the tubes are compared, namely a single filament in each tube and a scenario where the current in each tube is split into subfilaments. The latter scenario is shown to be the more realistic distribution because it leads to much better agreement with infrared camera observations. It is demonstrated that stripe patterns in the divertor heat flux produced by an ELM in the DIII-D tokamak can be reproduced numerically by taking into account the magnetic perturbation caused by the thermoelectric current subfilaments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2011

Accelerating the numerical simulation of magnetic field lines in tokamaks using the GPU

TRIP3D is a field line simulation code that numerically integrates a set of nonlinear magnetic field line differential equations. The code is used to study properties of magnetic islands and stochastic or chaotic field line topologies that are important for designing non-axisymmetric magnetic perturbation coils for controlling plasma instabilities in future machines. The code is very computationally intensive and for large runs can take on the order of days to complete on a traditional single CPU. This work describes how the code was converted from Fortran to C and then restructured to take advantage of GPU computing using NVIDIA's CUDA. The reduction in computing time has been dramatic where runs that previously took clays now take hours allowing a scale of problem to be examined that would previously not have been attempted. These gains have been accomplished without significant hardware expense. Performance, correctness, code flexibility, and implementation time have been analyzed to gauge the success and applicability of these methods when compared to the traditional multi-CPU approach.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2011

L-H transition studies on DIII-D to determine H-mode access for operational scenarios in ITER

A comprehensive set of L-H transition experiments has been performed on DIII-D to determine the requirements for access to H-mode plasmas in ITER's first (non-nuclear) operational phase with H and He plasmas and the second (activated) operational phase with D plasmas. The H-mode power threshold, P(TH), was evaluated for different operational configurations and auxiliary heating methods for the different main ion species. Helium plasmas have significantly higher P(TH) than deuterium plasmas at low densities for all heating schemes, but similar P(TH) as deuterium plasmas at high densities except for H-neutral beam injection-heated discharges, which are still higher. Changes in P(TH) are observed when helium concentration levels in deuterium plasmas exceed 40%. There is a strong dependence of P(TH) on the magnetic geometry in the vicinity of the divertor. The trend of decreasing P(TH) with decreasing X-point height is observed for all of the main ion species irrespective of the heating method, which appears to indicate that there is a common physics process behind this effect for all of the ion species. Helium and deuterium plasmas exhibit a significant increase in P(TH) for strong resonant magnetic perturbations. The application of a local magnetic ripple of 3% from test blanket module mock-up coils did not change P(TH) in deuterium plasmas.

Gohil, P↗

DOE OSTI2011

Effect of Applied Toroidal Electric Field on the Growth/Decay of Plateau-Phase Runaway Electron Currents in DIII-D

Large relativistic runaway electron currents (0.1-0.5 MA) persisting for similar to 100 ms are created in the DIII-D tokamak during rapid discharge shut down caused by argon pellet injection. Slow upward and downward ramps in runaway currents were found in response to externally applied loop voltages. Comparison between the observed current growth/decay rate and the rate expected from the knock-on avalanche mechanism suggests that classical collisional dissipation of runaways alone cannot account for the measured growth/damping rates. It appears that a fairly constant anomalous dissipation rate of order 10 s(-1) exists, possibly stemming from radial transport or direct orbit losses to the vessel walls, although the possibility of an apparent loss due to current profile shrinking cannot be ruled out at present.

Hollmann, E M↗

DOE OSTI2011

ITER Test Blanket Module Error Field Simulation Experiments at DIII-D

Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Delta upsilon/upsilon similar to 60% via non-resonant braking. Changes to global Delta n/n, Delta beta/beta and Delta H(98)/H(98) were similar to 3 times smaller. These effects are stronger at higher beta. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L-and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

Schaffer, M J↗

DOE OSTI2011

KSTAR Equilibrium Operating Space and Projected Stabilization at High Normalized Beta

Along with an expanded evaluation of the equilibrium operating space of the Korea Superconducting Tokamak Advanced Research, KSTAR, experimental equilibria of the most recent plasma discharges were reconstructed using the EFIT code. In near-circular plasmas created in 2009, equilibria reached a stored energy of 54 kJ with a maximum plasma current of 0.34 MA. Highly shaped plasmas with near double-null configuration in 2010 achieved H-mode with clear edge localized mode (ELM) activity, and transiently reached a stored energy of up to 257 kJ, elongation of 1.96 and normalized beta of 1.3. The plasma current reached 0.7 MA. Projecting active and passive stabilization of global MHD instabilities for operation above the ideal no-wall beta limit using the designed control hardware was also considered. Kinetic modification of the ideal MHD n = 1 stability criterion was computed by the MISK code on KSTAR theoretical equilibria with a plasma current of 2 MA, internal inductance of 0.7 and normalized beta of 4.0 with simple density, temperature and rotation profiles. The steep edge pressure gradient of this equilibrium resulted in the need for significant plasma toroidal rotation to allow thermal particle kinetic resonances to stabilize the resistive wall mode (RWM). The impact of various materials and electrical connections of the passive stabilizing plates on RWM growth rates was analysed, and copper plates reduced the RWM passive growth rate by a factor of 15 compared with stainless steel plates at a normalized beta of 4.4. Computations of active RWM control using the VALEN code showed that the n = 1 mode can be stabilized at normalized beta near the ideal wall limit via control fields produced by the midplane in-vessel control coils (IVCCs) with as low as 0.83kW control power using ideal control system assumptions. The ELM mitigation potential of the IVCC, examined by evaluating the vacuum island overlap created by resonant magnetic perturbations, was analysed using the TRIP3D code. Using a combination of all IVCCs with dominant n = 2 field and upper/lower coils in an even parity configuration, a Chirikov parameter near unity at normalized poloidal flux 0.83, an empirically determined condition for ELM mitigation in DIII-D, was generated in theoretical high-beta equilibria. Chirikov profile optimization was addressed in terms of coil parity and safety factor profile.

Park, Y S↗

DOE OSTI2011

Pellet interaction with runaway electrons

We describe results from recent experiments studying interaction of solid polystyrene pellets with a runaway electron current channel generated after cryogenic argon pellet rapid shutdown of DIII-D. Fast camera imaging shows the pellet trajectory and continuum emission from the subsequent explosion, with geometric calibration providing detailed explosion analysis and runaway energy. Electron cyclotron emission also occurs, associated with knock-on electrons broken free from the pellet by RE which then accelerate and runaway, and also with a short lived hot plasma blown off the pellet surface. In addition, we compare heating and explosion times from observations and a model of pellet heating and breakdown by runaway interaction. (C) 2011 Elsevier B.V. All rights reserved

43 PARTICLE ACCELERATORS↗

DOE OSTI2011

Plasma-surface interactions during tokamak disruptions and rapid shutdowns

Recent progress in understanding of disruptions and in developing methods to avoid disruption damage is presented. Nearly complete mitigation of conducted heat loads has been achieved with high-Z gas jet shutdown. The resulting local radiation heat flash melting in the main chamber might be a concern in ITER, especially with beryllium walls. During the current quench, significant vessel forces can occur due to halo currents I-halo; however, these are found to fall reliably below a boundary of (halo current fraction times halo current peaking factor) <0.7 both experimentally and numerically. Numerical simulations indicate that runaway electrons (REs) could cause serious damage to hard-to reach components in ITER, making their suppression a high priority. During the current quench, less than 20% of the density required for collisional suppression of REs appears to have been achieved. Collisional suppression of REs may have been achieved, however, in full-current RE beams with gas injection. (C) 2010 Elsevier B.V. All rights reserved.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2011

Toroidally Resolved Structure of Divertor Heat Flux in RMP H-mode Discharges on DIII-D

As shown on DIII-D edge localized modes (ELMs) can be either completely eliminated or mitigated with resonant magnetic perturbation (RMP) fields. Two infrared cameras, separated 105 degrees toroidally, were used to make simultaneous measurements of ELM heat loads with high frame rates. Without the RMP fields ELMs display a variety of different heat load dynamics and a range of toroidal variability that is characteristic of their 3D structure. Comparing radial averages there is no asymmetry between two toroidal locations. With RMP-mitigated ELMs, the variability in the radially averaged power loads is significantly reduced and toroidal asymmetries in power loads are introduced. In addition to RMP ELM suppression scenarios an RMP scenario with only very small ELMs and very good confinement has been achieved.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2011

The Influence of Three-dimensional Stochastic Magnetic Boundaries on Plasma Edge Transport and the Resulting Plasma Wall Interaction

The three-dimensional (3D) features of plasma edge profiles and wall interaction patterns induced by edge resonant magnetic perturbation fields (RMP) are discussed comparing TEXTOR and DIII-D. We show that the scrape-off layer (SOL) profiles and decay lengths depend the edge safety factor, the RMP base mode as well as on the plasma rotation during RMP application indicating modification of SOL transport by the 3D perturbation fields. This is compatible with channeling of particle and heat efflux along open perturbed field lines in the very edge of the plasma boundary into a completely re-arranged, helically striated 3D divertor footprint. The distribution of the measured divertor heat and particle fluxes at DIII-D match the vacuum modeled magnetic footprint topology in L-mode while in H-mode the striation width exceeds the modeled footprint width by 15-30%. This 3D structure of the measured heat and particle fluxes results in a new situation for the material erosion properties and initial quantification of the net-erosion within the 3D footprint shows in L-mode a 50% decrease of the chemical erosion yield and evidence for a comparably small 15-20% increase in physical sputtering. Extrapolation of these findings to ITER by vacuum modeling of the magnetic footprint for the actual ELM control coils shows a similar vacuum magnetic footprint topology as found at DIII-D during RMP ELM suppression. However, the open field lines escape the CFC covered ITER divertor area potentially transferring net-erosion characteristics from the CFC domain onto the Tungsten including so far unconsidered heat and particle loads on this sensitive material.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗