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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Effect of Deuterium Gas Puff On The Edge Plasma In NSTX

This paper describes a detailed examination of the effects of a relatively small pulsed deuterium gas puff on the edge plasma and edge turbulence in NSTX. This gas puff caused little or no change in the line-averaged plasma density or total stored energy, or in the edge density and electron temperature up to the time of the peak of the gas puff. The radial profile of the Dα light emission and the edge turbulence within this gas puff did not vary significantly over its rise and fall, implying that these gas puffs did not significantly perturb the local edge plasma or edge turbulence. These measurements are compared with modeling by DEGAS 2, UEDGE, and with simplified estimates for the expected effects of this gas puff.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Disruption avoidance in the SINP-Tokamak by means of electrode-biasing at the plasma edge

Control of plasma disruption by a biased edge electrode is reported in SINP-Tokamak. The features that characterize a plasma disruption are reduced with increasing bias potential. The disruption can be completely suppressed with the concomitant stabilization of observed MHD modes that are allegedly precursors of the disruption. An m = 3/n = 1 tearing mode, which apparently causes disruption can be stabilized when a negative biasing potential is applied near the edge. These changes in the disruptive behavior with edge biasing are hypothesized to be due to changes in the current density profile.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Faraday-effect polarimetry for current profile measurement in the tokamak plasma edge

Toroidal current profile measurements in the tokamak plasma edge are critical for fusion plasma physics research and model validation. A three-wave Faraday-effect polarimeter-interferometer with a sub-centimeter spatial resolution is proposed on the DIII-D tokamak to determine the edge current profile via Abel inversion. By using probe beams with 316 µm wavelength, a low-field-side, vertical-view, single-pass optical layout covering the plasma edge region (R = 2.15–2.27 m) is assessed. Measurements with no greater than 0.1° polarimetric systematic uncertainty, no greater than 0.01° polarimetric root-mean-square noise (1 kHz bandwidth), and a 0.8 cm radial chord spacing are considered feasible based on the achieved performance of existing systems using similar wavelengths on fusion devices. Furthermore, synthetic diagnostic calculations taking various factors into account, such as diagnostic uncertainty and quality of magnetic flux surfaces, find that the edge current profile can be determined with up to 0.12 MA/m 2 uncertainty, or about 10% of the peak current density in the pedestal of an investigated high-confinement plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Modeling transient edge plasma transport with dynamic recycling

The work presents numerical simulation studies of the role that dynamic plasma recycling on the main wall and divertor target surfaces plays in transient edge plasma transport phenomena, such as edge localized modes (ELMs). The studies are performed by coupling the edge plasma transport code UEDGE [Rognlien et al., J. Nucl. Mater. 196–198, 347 (1992)] and the wall reaction–diffusion transport code FACE [Smirnov et al., Fusion Sci. Technol. 71, 75 (2017)]. The two-dimensional, time-dependent, two-way coupling of the codes, in a realistic tokamak geometry, is accomplished using the Integrated Plasma Simulator framework [Elwasif et al., in 18th Euromicro Conference on Parallel, Distributed and Network-Based Processing (PDP 2010), Pisa, Italy (IEEE, 2010), pp. 419–427] for all modeled material plasma boundaries. The simulations show that dynamic plasma recycling has substantially different characteristics on the main wall and on the divertor plates. It is demonstrated that during an ELM cycle the outer wall can dynamically absorb and release a number of particles comparable to that expelled by the ELM from the core plasma, by far exceeding the dynamic retention capacity of the divertor surfaces. The resulting evolution of the edge and divertor plasma conditions during an ELM cycle is analyzed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Synthetic plasma edge diagnostics for EMC3-EIRENE, highlighted for Wendelstein 7-X

Interpretation of spectroscopic measurements in the edge region of high-temperature plasmas can be a challenge since line of sight integration effects make direct interpretation in terms of quantitative, local emission strengths often impossible. The EMC3-EIRENE code—a 3D fluid edge plasma and kinetic neutral gas transport code—is a suitable tool for full 3D reconstruction of such signals. A versatile synthetic diagnostic module has been developed recently which allows the realistic 3D setup of various plasma edge diagnostics to be captured. As a result, we highlight these capabilities with two examples for Wendelstein 7-X (W7-X): a visible camera for the analysis of recycling, and a coherent-imaging system for velocity measurements.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Divertor detachment characterization in negative triangularity discharges in DIII-D via 2D edge-plasma transport modeling

Edge fluid modeling of the first divertor-plasma detachment experiments in negative triangularity (NT) discharges in the DIII-D tokamak is presented using the 2D multifluid edge transport code UEDGE, including cross-field particle drifts. Experiments were performed where the lower single-null magnetic equilibrium had a strong NT (δ≈−0.5), that is, where the magnetic X-point is at a larger major radius than the core magnetic axis. Divertor-plasma detachment was induced by increasing the core plasma density in DIII-D via intrinsic gas puffing. Here density scans are performed with UEDGE to reach a detached plasma and to quantitatively recover the experimental rollover of the ion saturation current on the outer divertor target plate. The simulations cover experiments with both signs of the toroidal magnetic field, B T , where the ion magnetic Grad-B drifts are directed into (forward B T ) and out of (reverse B T ) the divertor region. Consistent with experiments with neutral beam power injection, the NT simulations reproduce: 40% higher density is needed to reach detachment onset with forward B T compared with reverse B T , and the absence of deep detachment is found with reverse B T . Similarly, comparison between Ohmic discharges in NT and positive triangularity (PT) shaping confirms that a substantially higher density is needed to achieve detachment in NT than in PT, with NT requiring an line-average density of at least the Greenwald density or higher. Simulation results suggest that higher densities are needed to reach detachment in negative compared to PT because these discharges have a shorter midplane-to-target distance along the total magnetic field B, a shorter outer divertor poloidal leg length (0.06 m vs 0.2 m), and reduced radial transport [near-scrape-off layer (SOL) D ⊥ /χ ⊥ =0.3/0.5 vs D ⊥ /χ ⊥ =1.0/1.0, all in (m 2 /s)].

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Impact of magnetic islands in the plasma edge on particle fueling and exhaust in the HSX and W7-X stellarators

The edge magnetic structure in the Helically Symmetric eXperiment (HSX) and Wendelstein 7X (W7-X) stellarators has been shown to have a significant impact on the particle fueling and exhaust of the plasma main species (hydrogen) as well as impurity helium. For HSX, the plasma sourcing to exhaust ratio, quantified by the effective and global particle confinement times τ p * and τ p , H , respectively, increases when a magnetic island chain is located in the plasma edge. The fueling efficiency is reduced by 25% when the plasma boundary is deformed by the magnetic islands. The X-point geometry also yields higher plasma temperatures in front of the main recycling region. When the island is moved radially inward, both τ p * and τ p decrease by 10 % – 25 % depending on plasma density. The τ p , H results rely heavily on EMC3-EIRENE modeling which confirms reduced fueling efficiency due to more rapid ionization in the outward shifted island position. These findings suggest that for a helically optimized system like HSX, the plasma fueling from the recycling source, as well as from active gas injection, can be controlled by the magnetic island chain in the plasma edge—which is a basic requirement for a divertor system. This process is also effective for the control of effective helium exhaust times, as τ p , H e * measured by perturbative gas puff experiments is reduced by up to 40% when the islands are shifted inwards. For Wendelstein 7-X, a similar reduction of τ p , H e * was inferred when magnetic islands were moved from the far plasma edge into the confined plasma region. Finally, however, the effective confinement features of H as the main plasma species were not affected due to the non-optimal position of the magnetic islands with respect to the highly localized ionization domain during the limiter startup campaign.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Gyrokinetic study of collisional resonant magnetic perturbation (RMP)-driven plasma density and heat transport in tokamak edge plasma using a magnetohydrodynamic screened RMP field

The total-f gyrokinetic particle-in-cell code XGC is applied to study various aspects of collisional transport in the tokamak edge pedestal in the presence of resonant magnetic perturbations (RMPs) calculated with M3D-C1. Simulations including the separatrix and scrape-off layer are exercised for a model DIII-D H-mode plasma. Neutral particle recycling is modeled by means of ionization and charge exchange with the plasma. A fully nonlinear Fokker–Planck collision operator is utilized to account for non-Maxwellian edge plasma. Here, the study yields kinetic evidence that (i) neglecting the non-axisymmetric component of the RMP-generated electrostatic potential perturbation could yield a fictitiously high particle pump-out rate in the edge pedestal; that (ii) the experimental level of particle pump-out from the pedestal cannot be provided by collisional transport across Kolmogorov–Arnold–Moser surfaces except across the magnetic separatrix where stochastic magnetic field lines exist; that (iii) the H-mode type electron heat barrier is retained in the steep slope area under the M3D-C1 calculated RMPs; and that (iv) the stochastic electron heat-transport rate around the magnetic separatrix surface is much lower than the prediction by Rechester–Rosenbluth (1978 Phys. Rev. Lett. 40 38–41) and is qualitatively consistent with the experimentally observed rate. Detailed kinetic physics analysis on how the transport fluxes, electrostatic potential perturbation, and -well structure respond to onset of RMPs is also reported.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Numerical investigation of plasma edge transport and limiter heat fluxes in Wendelstein 7-X startup plasmas with EMC3-EIRENE

Here the results of a first systematic assessment of plasma edge transport processes for the limiter startup configuration at Wendelstein 7-X are presented. This includes an investigation of transport from intrinsic and externally injected impurities and their impact on the power balance and limiter heat fluxes. The fully 3D coupled plasma fluid and kinetic neutral transport Monte Carlo code EMC3-EIRENE is used. The analysis of the magnetic topology shows that the poloidally and toroidally localized limiters cause a 3D helical scrape-off layer (SOL) consisting of magnetic flux tubes of three different connection lengths L C . The transport in the helical SOL is governed by L C as topological scale length for the parallel plasma loss channel to the limiters. A clear modulation of the plasma pressure with L C is seen. The helical flux tube topology results in counter streaming sonic plasma flows. The heterogeneous SOL plasma structure yields an uneven limiter heat load distribution with localized peaking. Assuming spatially constant anomalous transport coefficients, increasing plasma density yields a reduction of the maximum peak heat loads from 12 MWm -2 to 7.5 MWm -2 and a broadening of the deposited heat fluxes. The impact of impurities on the limiter heat loads is studied by assuming intrinsic carbon impurities eroded from the limiter surfaces with a gross chemical sputtering yield of 2 . The resulting radiative losses account for less than 10% of the input power in the power balance with marginal impact on the limiter heat loads. It is shown that a significant mitigation of peak heat loads, 40–50%, can be achieved with controlled impurity seeding with nitrogen and neon, which is a method of particular interest for the later island divertor phase.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2018

On the role of finite grid extent in SOLPS-ITER edge plasma simulations for JET H-mode discharges with metallic wall

The impact of the finite grid size in SOLPS-ITER edge plasma simulations is assessed for JET H-mode discharges with a metal wall. For a semi-horizontal divertor configuration it is shown that the separatrix density is at least 30% higher when a narrow scrape-off layer (SOL) grid width is chosen in SOLPS-ITER compared to the case for which the SOL grid width is maximised. The density increase is caused by kinetic neutrals being not confined inside the divertor region because of the reduced extent of the plasma grid. In this case, an enhanced level of reflections of energetic neutrals at the low-field side (LFS) metal divertor wall is observed. This leads to a shift of the ionisation source further upstream which must be accounted for as a numerical artefact. An overestimate in the cooling at the divertor entrance is observed in this case, identified by a reduced heat flux decay parameters λ q div . Otherwise and further upstream the mid-plane heat decay length λ q parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that λ q upstream is essentially set by the assumptions for the ratio radial to parallel heat conductivity. It is also shown that even for attached conditions the decay length relations λ ne > λ Te > λ q hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of λ ne into account, rather than λ q , as the former actually defines the required minimum upstream SOL grid extent.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Accuracy and convergence of coupled finite-volume/Monte Carlo codes for plasma edge simulations of nuclear fusion reactors

The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracy by making use of averaging in the Random Noise coupling technique.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2023

Modeling of carbon and tungsten transient dust influx in tokamak edge plasma

The paper presents computer simulation studies of burst injection of carbon and tungsten dust particles in DIII-D-like edge plasmas. The injection causes a large transient influx of the low- and high-Z impurities associated with the dust ablation in the plasmas. The dust transport and the effects of the ablated impurities on the edge plasma dynamics in a modern mid-size tokamak geometry are investigated for low- and high-power plasma discharge conditions. The core plasma contamination with dust-ablated impurities and the factors affecting it are evaluated.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Electromagnetic turbulence simulation of tokamak edge plasma dynamics and divertor heat load during thermal quench

Abstract The edge plasma turbulence and transport dynamics, as well as the divertor power loads during the thermal-quench phase of tokamak disruptions, are numerically investigated with BOUT++’s flux-driven six-field electromagnetic turbulence model. Here, transient yet intense particle and energy sources are applied at the pedestal top to mimic the plasma power drive at the edge induced by a core thermal collapse, which flattens the core temperature profile. Interesting features, such as surging of divertor heat load (up to 50 times) and broadening of heat-flux width (up to four times) on the outer-divertor target plate, are observed in the simulation, in qualitative agreement with experimental observations. The dramatic changes in divertor heat load and width are due to the enhanced plasma turbulence activities inside the separatrix. Two cross-field transport mechanisms, namely, the E × B turbulent convection and the stochastic parallel advection/conduction, are identified to play important roles in this process. First, an elevated edge pressure gradient drives instabilities and subsequent turbulence in the entire pedestal region. The enhanced turbulence not only transports particles and energy radially across the separatrix via the E × B convection, which causes the initial divertor heat-load burst, but it also induces amplified magnetic fluctuation B ˜ . Once themagnetic fluctuation is large enough to break the magnetic flux surface, magnetic flutter effect provides an additional radial transport channel. In the late stage of our simulation, | B ˜ r / B 0 | reaches to 10 −4 level that completely breaks magnetic flux surfaces such that stochastic field lines are directly connecting pedestal top plasma to the divertor target plates or first wall, further contributing to the divertor heat-flux width broadening.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

SciDAC - Center for Plasma Edge Simulation - Project Summary

Final Technical Report: Center for Plasma Edge Simulation (CPES) Principal Investigator: Scott Parker, University of Colorado, Boulder Description/Abstract First-principle simulations of edge pedestal micro-turbulence are performed with the global gyrokinetic turbulence code GEM for both low and high confinement tokamak plasmas. The high confinement plasmas show a larger growth rate, but nonlinearly a lower particle and heat flux. Numerical profiles are obtained from the XGC0 neoclassical code. XGC0/GEM code coupling is implemented under the EFFIS (“End-to-end Framework for Fusion Integrated Simulation”) framework. Investigations are underway to clearly identify the micro-instabilities in the edge pedestal using global and flux-tube gyrokinetic simulation with realistic experimental high confinement profiles. We use both experimental profiles and those obtained using the EFFIS XGC0/GEM coupled code framework. We find there are three types of instabilities at the edge: a low-n, high frequency electron mode, a high-n, low frequency ion mode, and possibly an ion mode like kinetic ballooning mode (KBM). Investigations are under way for the effects of the radial electric field. Finally, we have been investigating how plasmas dominated by ion-temperature gradient (ITG) driven turbulence, how cold Deuterium and Tritium ions near the edge will naturally pinch radially inward towards the core. We call this mechanism “natural fueling.” It is due to the quasi-neutral heat flux dominated nature of the turbulence and still applies when trapped and passing kinetic electron effects are included. To understand this mechanism, examine the situation where the electrons are adiabatic, and there is an ion heat flux. In such a case, lower energy particles move inward and higher energy particles move outward. If a trace amount of cold particles are added, they will move inward.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Lagrangian conditional statistics and flow topology in edge plasma turbulence

Lagrangian statistics and particle transport in edge plasma turbulence are investigated using the Hasegawa–Wakatani model and its modified version. The latter shows the emergence of pronounced zonal flows. Different values of the adiabaticity parameter are considered. The main goal is to characterize the role of coherent structures, i.e., vortices and zonal flows, and their impact on the Lagrangian statistics of particles. Computationally intensive long time simulations following ensembles of test particles over hundreds of eddy turnover times are considered in statistically stationary turbulent flows. The flow topology is characterized using the Lagrangian Okubo–Weiss criterion in order to split the flow into topologically different domains. In elliptic and hyperbolic regions, the probability density functions (PDFs) of the residence time have self-similar algebraic decaying tails. However, in the intermediate regions, the PDFs exhibit exponentially decaying tails. Topologically conditioned PDFs of the Lagrangian velocity, and acceleration and density fluctuations are likewise computed. The differences between the classical Hasegawa–Wakatani system and its modified version are assessed, and the role of zonal flows is highlighted. Here, the density flux spectrum, which characterizes the contributions of different length scales, is studied, and its inertial scaling is found to be in agreement with predictions based on dimensional arguments. Analyzing the angular change of particle tracers at different time scales, corresponding to coarse grained curvature, completes the study, and these multiscale geometric statistics quantify the directional properties of the particle motion in different flow regimes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

The three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

Results from three-dimensional modeling of plasma edge transport and plasma–wall interactions during application of resonant magnetic perturbation (RMP) fields for control of edge-localized modes in the ITER standard 15 MA Q = 10 H-mode are presented. The full 3D plasma fluid and kinetic neutral transport code EMC3-EIRENE is used for the modeling. Four characteristic perturbed magnetic topologies are considered and discussed with reference to the axisymmetric case without RMP fields. Two perturbation field amplitudes at full and half of the ITER ELM control coil current capability using the vacuum approximation are compared to a case including a strongly screening plasma response. In addition, a vacuum field case at high q95 = 4.2 featuring increased magnetic shear has been modeled. Formation of a three-dimensional plasma boundary is seen for all four perturbed magnetic topologies. The resonant field amplitudes and the effective radial magnetic field at the separatrix define the shape and extension of the 3D plasma boundary. Opening of the magnetic field lines from inside the separatrix establishes scrape-off layer-like channels of direct parallel particle and heat flux towards the divertor yielding a reduction of the main plasma thermal and particle confinement. This impact on confinement is most accentuated at full RMP current and is strongly reduced when screened RMP fields are considered, as well as for the reduced coil current cases. The divertor fluxes are redirected into a three-dimensional pattern of helical magnetic footprints on the divertor target tiles. At maximum perturbation strength, these fingers stretch out as far as 60 cm across the divertor targets, yielding heat flux spreading and the reduction of peak heat fluxes by 30%. However, at the same time substantial and highly localized heat fluxes reach divertor areas well outside of the axisymmetric heat flux decay profile. Reduced RMP amplitudes due to screening or reduced RMP coil current yield a reduction of the width of the divertor flux spreading to about 20–25 cm and cause increased peak heat fluxes back to values similar to those in the axisymmetric case. The dependencies of these features on the divertor recycling regime and the perpendicular transport assumptions, as well as toroidal averaged effects mimicking rotation of the RMP field, are discussed in the paper.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Modelling of edge plasma dynamics with active wall boundary conditions

A self-consistent 2D model is presented for transport in boundary plasma and plasma-facing material walls. Plasma dynamics in the domain is represented by a 2D collisional plasma fluid model in the edge-plasma code UEDGE, and transport of hydrogen and heat in the wall is represented by a system of reaction–diffusion equations in the 1D wall code FACE. To account for variation of parameters along the wall, in the coupled model multiple instances of the FACE code run in parallel. Here, the coupled model provides a tool for investigating a range of dynamic plasma–material interactions phenomena in 2D. For demonstration of its capability, one application of particular interest is the role of active wall in tokamak strike point sweeping proposed for mitigation of divertor heat loads. In the present study, the coupled calculations are applied to investigation of the impact of heat and hydrogen transport in the material wall on the divertor plasma and target heat load during sweeping of the target strike point for parameters of a high-power tokamak.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Temporal parallelization of edge plasma simulations using the parareal algorithm and the SOLPS code

We show that numerical modelling of edge plasma physics may be successfully parallelized in time. The parareal algorithm has been employed for this purpose and the SOLPS code package coupling the B2.5 finite-volume fluid plasma solver with the kinetic Monte-Carlo neutral code Eirene has been used as a test bed. The complex dynamics of the plasma and neutrals in the scrape-off layer (SOL) region makes this a unique application. It is demonstrated that a significant computational gain (more than an order of magnitude) may be obtained with this technique. The use of the IPS framework for event-based parareal implementation optimizes resource utilization and has been shown to significantly contribute to the computational gain.

97 MATHEMATICS AND COMPUTING↗