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Results for “plasma-wall interaction”

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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33 records · Page 2
All sourcesNASA NTRSDOE OSTI
NASA NTRS2019

Variation in Ion Acceleration Characteristics of the HERMeS Hall Thruster During Magnetic Optimization

To reduce design risks for future magnetically shielded Hall thrusters, a test was performed on the HERMeS to obtain data for optimizing the effect of magnetic shielding. As a part of this test, laser-induced fluorescence velocimetry was used to characterize the variations in the ion acceleration with different magnetic configurations. Four magnetic configurations representing varying amounts of magnetic shielding between the high-energy discharge plasma and the discharge channel walls were tested. The ion velocity data points to the possibility that different plasma-wall interaction physics applies to a magnetically shielded thruster than a non-shielded thruster. The transition point is very prominent and can potentially be used to test whether a thruster is fully magnetically shielded.

HERMeS↗
DOE OSTI2021

2D radial-azimuthal particle-in-cell benchmark for E × B discharges

In this paper we propose a representative simulation test-case of E × B discharges accounting for plasma wall interactions with the presence of both the electron cyclotron drift instability and the modified-two-stream-instability. Seven independently developed particle-in-cell (PIC) codes have simulated this benchmark case, with the same specified conditions. The characteristics of the different codes and computing times are given. Here, results show that both instabilities were captured in a similar fashion and good agreement between the different PIC codes is reported as main plasma parameters were closely related within a 5% interval. The number of macroparticles per cell was also varied and statistical convergence was reached. Detailed outputs are given in the supplementary data, to be used by other similar groups in the perspective of code verification.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Anticathode effect on electron kinetics in electron beam generated E × B plasma

Electron beam (e-beam) generated plasmas with applied crossed electric and magnetic (E x B) fields are promising for low damage processing of materials with applications to microelectronics and quantum information systems. In cylindrical e-beam E x B plasmas, radial confinement of electrons and ions is achieved by an axial magnetic field and radial electric field, respectively. To control the axial confinement of electrons, such e-beam generated plasma sources may incorporate a conducting boundary known as an anticathode, which is placed on the axially opposite side of the plasma from the cathode. In this work, it is shown that varying the anticathode voltage bias can control the degree to which the anticathode collects or repels incident electrons, allowing control of warm electron (electron energies in 10–30 eV range) and beam electron population confinement. It is suggested that the effect of the anticathode bias on the formation of these distinct electron populations is also associated with the transition between weak turbulence and strong Langmuir turbulence.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Interpretive modelling of boron transport in the boundary plasma of WEST experiments with the impurity powder dropper

Abstract Boron (B) powder injection is a potential alternative to glow discharge boronization as a wall conditioning method for tokamaks. This technique is currently being studied in WEST experiments, during which B powder is injected by an Impurity Powder Dropper developed by PPPL. In order to interpret and analyse experimental trends, and to help develop future experiments, a modelling workflow using a boundary plasma simulation (SOLEDGE-EIRENE) and powder ablation simulation (Dust Injection Simulator) was developed and tested. The effect of adding a B neutral source to simulated deuterium + oxygen (D + O) plasmas was compared to experimental data from the WEST C5 campaign, where B powder was injected in a dedicated experiment. While the impact of B injection on radiated power P rad measurements at the upper divertor was similar, there were significant differences in measurements of P rad , outer strike point electron temperature T e OSP and O-II line intensity at the lower divertor between experiment and simulation. This discrepancy suggests that those parameters were affected by phenomena not present in the simulations, with the most likely candidates being reduced D recycling and a reduced O sourcing from the divertor.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Small resonant magnetic perturbations result in three-dimensional material transport in the fusion plasma edge

Erosion patterns in a 3D magnetic topology are significantly altered when compared to an axisymmetric scenario due to differences in the impurity transport in a plasma environment. When resonant magnetic perturbations are applied in L-Mode discharges, eroded impurities can buildup in regions where an axisymmetric plasma would otherwise lead to zones of net erosion across the full toroidal length of the divertor. The reduction on the local and integrated erosion observed across a parametric study of the anomalous diffusion and impurity content in a plasma, can lead to the extension of divertor lifetimes in low power scenarios when 3D fields are applied. By contrast, in axisymmetric scenarios, most of the carbon eroded from the divertor are carried away to the far scrape-off layer.

DIII-D↗
DOE OSTI2025

Mitigation of ICRF—edge plasma interaction in Alcator C-Mod

The parasitic interaction between ion cyclotron range of frequencies (ICRF) waves and the edge plasma has long limited high-power ICRF utilization in magnetically confined fusion plasmas. In this paper, we report on the successful mitigation of this interaction by power tapering a four-strap field-aligned (FA) antenna in the Alcator C-Mod tokamak. With the antenna operated in dipole phasing ($0\pi0\pi$), the ratio of the power coupled by the central two straps ($P_\mathrm {cent}$) to the power coupled by the outer two straps was varied at fixed total coupled ICRF power ($P_\mathrm {tot}$). With $P_\mathrm {cent}$/$P_\mathrm {tot} \approx 0.8-0.9$, no enhancement of the plasma potential was measured despite 1 MW of coupled ICRF power. When the antenna is powered, energy deposition increases on its four corners and on neighboring limiter tiles magnetically connected to some of those corners. The energy deposited on the antenna and limiter is minimized together with near-field plasma potentials. Additionally, we show that deviation from $P_\mathrm {cent}$/$P_\mathrm {tot} \approx 0.8$ enhances wave-plasma interactions in places not magnetically connected to the antenna. These far-field effects are correlated with changes in the antenna spectrum. Hence, on C-Mod, near-field and far-field interactions were minimized for about the same range of $P_\mathrm {cent}$/$P_\mathrm {tot}$. Optimal power tapering was also found to improve ICRF heating efficiency and facilitate H-mode access. Finally, we show that the heat flux pattern on the toroidally aligned antennas reverses with the direction of the RF-induced E×B flow in front of the antenna, a correlation not observed for the FA antenna. We argue that the RF-induced E×B flow, which carries high-density plasma from the near-SOL, is aligned with the FA antenna and does not intercept its corners, where the plasma potential is large. This could explain the previously reported smaller impurity generation by the FA antenna (Wukitch et al 2013 Phys. Plasmas 20 056117).

Alcator C-Mod↗
DOE OSTI2026

Results from the last DD and DT JET campaigns in the framework of the EUROfusion Tokamak Exploitation Work Package activity

JET, the only tokamak capable of operating with deuterium–tritium (D–T) fuel (since TFTR was shutdown in 1999), has provided essential experimental data to support ITER and DEMO design and operation. Within the EUROfusion Tokamak Exploitation Work Package, JET completed its final campaigns (2022–2023), culminating in the third D–T campaign (DTE3). These experiments addressed key challenges in plasma scenarios, exhaust control, and tritium management under reactor-relevant conditions. Significant progress was achieved in demonstrating ITER-like integrated scenarios with impurity seeding, achieving partial divertor detachment and high confinement ($H_{98}(y,2)$ ≈ 0.85) at 3 MA in D–T plasmas. Advanced exhaust regimes such as quasi-continuous exhaust (QCE) and X-point radiator (XPR) were successfully achieved first in D–D and then extended to D–T operation, confirming their relevance for mixed isotope operation. Operational milestones included a new world record of 69 MJ fusion energy in tritium-rich hybrid plasmas and long-pulse H-mode operation up to 60 s, contributing with unique data to the CICLOP database. Physics studies focused on peeling-limited pedestals in support of ITER and improved understanding of edge stability and impurity screening in metallic environments. Extensive usage of the shattered pellet injector (SPI) on JET provided critical information for the design of the ITER disruption mitigation system (DMS). Real-time control systems for D/T ratio control and plasma exhaust were deployed and demonstrated in D–D and D–T, while energetic particle physics investigations unfolded the role of fast ions in turbulence suppression mechanisms. Comprehensive tritium retention studies using gas balance method, post-mortem analysis, and ITER-relevant laser induced desorption spectroscopy (LIDS) diagnostics provided essential input for tritium accountancy strategies. These results are validating the ITER operational concepts, inform DEMO design, and deliver critical experience in nuclear operation and scenario integration.

disruptions↗
DOE OSTI2020

Detachment in Fusion Plasmas with Symmetry Breaking Magnetic Perturbation Fields

Power exhaust from the bulk plasma is significantly altered by symmetry breaking magnetic perturbation fields, because these create direct connections (perturbed field lines) from the confined high temperature plasma to solid surfaces. Additionally, the same amount of power is distributed among those new exhaust channels as for a symmetric magnetic configuration, which reduces the local upstream heat flux flowing down the perturbed field lines, thereby making access to detachment easier (i.e., at lower upstream density) for the divertor plasma near the location corresponding to the symmetric magnetic separatrix. However, the divertor plasma regions with connection to the bulk plasma are extended nonaxisymmetrically further outside, where significant heat loads occur, unlike in the symmetric configuration. The temperature remains high at those locations, which reduces the divertor plasma dissipation capacity, making the mitigation of heat loads more difficult to achieve.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Bohm Criterion of Plasma Sheaths away from Asymptotic Limits

The plasma exit flow speed at the sheath entrance is constrained by the Bohm criterion. The so-called Bohm speed regulates the plasma particle and power exhaust fluxes to the wall, and it is commonly deployed as a boundary condition to exclude the sheath region in quasineutral plasma modeling. In this work, the Bohm criterion analysis is performed in the intermediate plasma regime away from the previously known limiting cases of adiabatic laws and the asymptotic limit of infinitesimal Debye length in a finite-size system, using the transport equations of an anisotropic plasma. The resulting Bohm speed has explicit dependence on local plasma heat flux, temperature isotropization, and thermal force. Comparison with kinetic simulations demonstrates its accuracy over the plasma-sheath transition region in which quasineutrality is weakly perturbed and the Bohm criterion applies.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Broadening of the Divertor Heat Flux Profile in High Confinement Tokamak Fusion Plasmas with Edge Pedestals Limited by Turbulence in DIII-D

Multimachine empirical scaling predicts an extremely narrow heat exhaust layer in future high magnetic field tokamaks, producing high power densities that require mitigation. Here, in the experiments presented, the width of this exhaust layer is nearly doubled using actuators to increase turbulent transport in the plasma edge. This is achieved in low collisionality, high confinement edge pedestals with their gradients limited by turbulent transport instead of large-scale, coherent instabilities. The exhaust heat flux profile width and divertor leg diffusive spreading both double as a high frequency band of turbulent fluctuations propagating in the electron diamagnetic direction doubles in amplitude. The results are quantitatively reproduced in electromagnetic XGC particle-in-cell simulations which show the heat flux carried by electrons emerges to broaden the heat flux profile, directly supported by Langmuir probe measurements.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Combined Influence of Rotation and Scrape-Off Layer Drifts on Recycling Asymmetries in Tokamak Plasmas

Coupled 2D fluid-kinetic simulations of a DIII-D high confinement tokamak plasma show that plasma rotation coupled with drift effects near the plasma edge play a significant role in the creation of the observed poloidal distribution of neutrals. It is observed that including either drift or rotation effects enhances particle flux at the inner target in the case of ion 𝐵×∇𝐵 drift toward the 𝑋-point. However, the particle flux asymmetry is significantly higher with the combination of drifts and rotation than either effect alone. The heightened particle flux asymmetry allows for improved simulation of the strong in-out asymmetry of the Lyman-𝛼 brightness profiles measured in the experiment. Enhancement of radial transport of parallel momentum changes the upstream scrape-off layer flow pattern, increasing the fraction of deuterium flux that reaches the inboard divertor entrance while lowering that which arrives at the outboard. In conclusion, this Letter indicates that by combining drifts, rotation, and viscous coupling, existing boundary plasma models can achieve a satisfactory agreement with experimentally measured neutral asymmetries.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Ripple-induced neoclassical toroidal viscous torque in Augmented-First Plasma operation phase in ITER

Abstract A systematic calculation is performed on the ripple-induced neoclassical toroidal viscous (NTV) torque for new ITER scenarios designed for the Augmented-First Plasma (A-FP) operation phase with the full tungsten wall, where the plasma-wall gap is varied in view of mitigating the impact of tungsten wall-plasma interactions. The torque calculation includes drift kinetic response of the plasma thermal and energetic particles to the n = 18 ( n is the toroidal harmonic number) ripple field. For the plasma scenario with ~45 cm plasma-wall gap at the outboard mid-plane and considering the corrected ripple level of 0.17% by the ferritic steel inserts, the computed net NTV torque acting on the plasma column is in the sub-Nm level. However, with decreasing the plasma-wall gap, the computed net NTV torque can reach a level comparable to that produced by the neutral-beam momentum injection in ITER. Ripple correction by ferritic inserts reduces the net torque by a factor of 3.3 for all the three A-FP scenarios considered. The n ω d = l ω b (with ω d and ω b being the toroidal precession and bounce frequencies of trapped particles, respectively, and l an integer number) type of resonance-enhancement of the NTV torque, due to thermal particles, is found to be weak in ITER despite high- n of 18. The same also holds for the ITER 10 MA steady state scenario from the D-T operation phase, where the aforementioned resonance associated with fusion-born alphas is also included. The ripple-induced NTV torque is well below that produced by the resonant magnetic perturbation applied for controlling the type-I edge-localized mode in ITER.

ITER↗
DOE OSTI2023

Validation of the plasma-wall self-organization model for density limit in ECRH-assisted start-up of Ohmic discharges on J-TEXT

Abstract A recently developed plasma-wall self-organization model (Escande et al 2022 Nucl. Fusion 62 026001) predicts a significantly enhanced density limit, which may be attainable in tokamaks with ECRH-assisted Ohmic start-up and sufficiently high initial neutral density. Experiments have been conducted on J-TEXT to validate such a density limit scenario based on this model. Experimental results demonstrate that increasing the pre-filled gas pressure or ECRH power during the start-up phase can help effectively enhance plasma purity and raise the density limit at the flat-top phase. The experimental data shows that density limit increases by a ratio 0.76 / 0.6 ≈ 1.27 for a decrease by a factor 0.13 / 0.33 ≈ 0.4 in C III radiation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Stability Boundaries of Discharges Generated in the Focal Region of a CW Microwave Beam

Plasma was generated in the focal region of a multikilowatt, 4.7-GHz continuous-wave (CW) microwave beam far removed from the walls of the vacuum chamber, minimizing the nonideal effects associated with plasma-wall and beam-wall interactions. Experiments were conducted with gas mixtures of argon, nitrogen, and oxygen at gas pressures ranging from 100 to 200 mtorr. Furthermore, three types of discharges were observed during these experiments: unstable, quasi-stable, and stable discharges. It was determined that the stability of the discharges could be controlled through adjustments of the gas composition, gas pressure, and power of the microwave beam. This article reports on the operational conditions in which stable discharges were achieved and the general behavior of these discharges as the control parameters were varied.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS2003

Magnetic Flux Compression Experiments Using Plasma Armatures

Magnetic flux compression reaction chambers offer considerable promise for controlling the plasma flow associated with various micronuclear/chemical pulse propulsion and power schemes, primarily because they avoid thermalization with wall structures and permit multicycle operation modes. The major physical effects of concern are the diffusion of magnetic flux into the rapidly expanding plasma cloud and the development of Rayleigh-Taylor instabilities at the plasma surface, both of which can severely degrade reactor efficiency and lead to plasma-wall impact. A physical parameter of critical importance to these underlying magnetohydrodynamic (MHD) processes is the magnetic Reynolds number (R(sub m), the value of which depends upon the product of plasma electrical conductivity and velocity. Efficient flux compression requires R(sub m) less than 1, and a thorough understanding of MHD phenomena at high magnetic Reynolds numbers is essential to the reliable design and operation of practical reactors. As a means of improving this understanding, a simplified laboratory experiment has been constructed in which the plasma jet ejected from an ablative pulse plasma gun is used to investigate plasma armature interaction with magnetic fields. As a prelude to intensive study, exploratory experiments were carried out to quantify the magnetic Reynolds number characteristics of the plasma jet source. Jet velocity was deduced from time-of-flight measurements using optical probes, and electrical conductivity was measured using an inductive probing technique. Using air at 27-inHg vacuum, measured velocities approached 4.5 km/s and measured conductivities were in the range of 30 to 40 kS/m.

Turner, M. W.↗