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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Fast response, optical radiance measurements of low intensity impurity emission in WEST plasmas with staggered wavelength filters

Fast response acquisition of atomic line emission through the use of a filterscope has been implemented on the WEST tokamak at multiple poloidal locations. Filterscopes consist of a fiber-optic-based transmission from the tokamak to specifically engineered optical bandpasses and photomultiplier tubes (PMT) that collect the emission intensity to measure a radiance, L [W/cm2/str], over the specified bandpass. This diagnostic records plasma-wall interaction (PWI) properties (e.g. impurity emission and recycling of main fuel ions) up to a max acquisition rate of 100 kHz. These PMT radiometric measurements are calibrated into a line-normalized radiance, LN [photons/sec/cm2/str], similar to traditional spectrometers, which can later be converted to a particle flux. Low intensity emission peaks are difficult to quantify due to often comparable continuum levels, thus a secondary filter shifted to a judiciously-selected, line-free region (~1 nm away) allows for background subtraction. The system currently installed on WEST targets tungsten (W) gross sputtering specifically by monitoring, with a filter pair, the spectral region near the neutral W line emission (W I 400.9 nm). The line and background filters of the pair are centered at 400.6 nm and 403.1 nm, respectively. Through a set beam splitters on each sight line, the filter pair are measuring from the same location. The two L signals near W I are then subtracted from each other yielding only the W I LN. W I LN data from a recent WEST experimental campaign is presented and compared with plasma parameters and traditional spectrometer measurements of W I line emission to demonstrate the capabilities of this staggered-filter filterscope method. Specifically, the choice of WEST-specific bandpass curves for the W I LN will be illustrated based on this latter comparison.

Neff, Anton↗
DOE OSTI2022

The radiated power limit in impurity seeded JET-ILW plasmas

The total radiated fraction is examined in high density H-mode plasmas (Greenwald fraction of about 85 %) in JET by the variation of the auxiliary heating power of Pheat = 14 MW-29 MW. An achieved radiation fraction of about 75 % at most has been observed in JET-ILW, which is less than the highest achievable (≈90 %) fraction in JET-C during the high radiative power scenarios with N2 seeding. It is shown that the maximal achievable total radiation fraction averaged over ELM cycles has a strong dependence on the radiation efficiency of the ELM energy, $θ_{rad}$: $\gamma^{max}_{rad,total}$ = 1 -$\frac {f_{ELM} x ΔW_{ELM}} {P_heat}$ (1- $θ_{rad}$). About 50 % and 16 % of the ELM induced diamagnetic energy drop (ΔW ELM ) radiates during the ELM in JET-C and JET-ILW, respectively, which corresponds to the maximum total radiated powers of $\gamma^{max}_{rad,JET-C}$ = 0.87 and $\gamma^{max}_{rad,JET-ILW}$ = 0.77. These values of the maximum of the radiative power fractions are in good agreement with $\gamma^{max}_{rad}$ experimentally observed in JET-C (90 %) and JET-ILW (75 %).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

A tutorial on the micro-trench technique for incident ion angle, material erosion, and impurity deposition measurements at plasma-facing surfaces

The micro-trench method is a new technique to measure polar and azimuthal incident ion angles, material erosion, and impurity deposition at plasma-facing surfaces, and has been applied in the DIII-D divertor. This article gives a tutorial of the micro-trench technique consisting of six steps: (1) micro-trench fabrication, (2) pre-exposure measurement of the fabricated micro-trench geometry, (3) tracer material deposition, (4) plasma exposure, (5) post-exposure observation of the tracer material, and (6) post-exposure measurement of the micro-trench geometry. Two criteria need to be satisfied to apply the micro-trench method successfully: (i) uniform impurity deposition on the micro-trench floor, and (ii) erosion dominated by the physical sputtering induced by the impinging ions. Furthermore, when those two criteria are satisfied, post-exposure analysis of the impurity deposition patterns on the micro-trench floor may be used to determine the polar and azimuthal incident ion directions (mean values of the ion angle directions), erosion rate, and impurity deposition rate during plasma exposure without computational interpretation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Characterization and controllability of radiated power via extrinsic impurity seeding in strongly negative triangularity plasmas in DIII-D

Experiments with extrinsic impurity seeding in strongly negative triangularity shapes in DIII-D achieved radiated power fractions (relative to input power) of up to ≈85% total radiation and ≈55% core radiation in steady-operating conditions. The relationship between core and total radiation was sensitive to impurity species and input power. Attempts to reach higher radiation levels via higher impurity flows resulted in radiative collapse disruptions. Nitrogen, neon, argon, and krypton were tested. Injection was by gas puffing, usually controlled by feeding back real-time estimates for core or total radiating fractions ($P_\textrm{rad} / P_\textrm{input}$). Argon and krypton were controlled by feeding back total radiating fraction, whereas neon was controlled by feeding back the core radiating fraction, due to the lower efficiency of neon as a divertor radiator. Nitrogen flows were pre-programmed. Reasonable total $P_\textrm{rad}$ control target following was achieved with argon or krypton. Control was more challenging with the neon/core radiation configuration, which was more prone to slow response and overshooting of the control target. Poor particle removal contributed to the control challenge: neon particle inventory within the last closed flux surface was roughly constant for up to 1 s (the longest duration tested) after neon injection was halted. However, separate experiments with laser-blow off of non-recycling impurities measured a short impurity confinement time, on the scale of the energy confinement time of ∼100 ms. Modeling with the Aurora impurity transport simulation matched experimental neon density profiles with full recycling (R = 1.0) and weak pumping but predicted rapid decreases in neon inventory if pumping were increased or recycling decreased. This indicates that changes outside the confined plasma (adding a well-placed pump) would improve controllability for all highly recycling species.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Experimental study on boron distribution and transport at plasma-facing components during impurity powder dropping in the Large Helical Device

Toward real-time wall conditioning, impurity powder dropping experiments with boron powder were performed in the 22nd experimental campaign of the Large Helical Device. To examine the deposition and desorption process of boron, we focus on boron hydride (BH) molecules which presumably populate near plasma-facing components. We performed spatially-resolved spectroscopic measurements of emission by boron ions and BH molecules. From the measurement, we found that BH and B + were concentrated on the divertor viewing chord, which suggest boron deposition in the divertor region. By comparing Hγ emissions with and without boron injection, neutral hydrogen shows uniform reduction in the SOL region, whereas less reduction of neutral hydrogen is confirmed in the divertor region. Although emissions from BH and B + increased linearly, emissions by B0 and B4 + became constant after the middle of the discharge. Continuous reduction of carbon density in the core plasma was confirmed even after B 0 and B 4 + became constant. The results may show reduction of hydrogen recycling and facilitation of impurity gettering by boron in the divertor region and thus effective real-time wall conditioning.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Turbulent impurity transport simulations in Wendelstein 7-X plasmas

A study of turbulent impurity transport by means of quasilinear and nonlinear gyrokinetic simulations is presented for Wendelstein 7-X (W7-X). The calculations have been carried out with the recently developed gyrokinetic code stella. Different impurity species are considered in the presence of various types of background instabilities: ion temperature gradient (ITG), trapped electron mode (TEM) and electron temperature gradient (ETG) modes for the quasilinear part of the work; ITG and TEM for the nonlinear results. While the quasilinear approach allows one to draw qualitative conclusions about the sign or relative importance of the various contributions to the flux, the nonlinear simulations quantitatively determine the size of the turbulent flux and check the extent to which the quasilinear conclusions hold. Although the bulk of the nonlinear simulations are performed at trace impurity concentration, nonlinear simulations are also carried out at realistic effective charge values, in order to know to what degree the conclusions based on the simulations performed for trace impurities can be extrapolated to realistic impurity concentrations. The presented results conclude that the turbulent radial impurity transport in W7-X is mainly dominated by ordinary diffusion, which is close to that measured during the recent W7-X experimental campaigns. Finally, it is also confirmed that thermodiffusion adds a weak inward flux contribution and that, in the absence of impurity temperature and density gradients, ITG- and TEM-driven turbulence push the impurities inwards and outwards, respectively.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Impact of a minority relativistic electron tail interacting with a thermal plasma containing high-atomic-number impurities

A minority relativistic electron component can arise in both laboratory and naturally-occurring plasmas. In the presence of high-atomic-number ion species, the ion charge state distribution at low bulk electron temperature can be dominated by relativistic electrons, even though their density is orders of magnitude lower. This is due to the relativistic enhancement of the collisional excitation and ionization cross sections. The resulting charge state effect can dramatically impact the radiative power loss rate and the related Bethe stopping power of relativistic electrons in a dilute plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Research and Development to Reduce Impurity Production and Transport of the Impurities to the Target in Linear Plasma Devices Using Helicon Plasma Sources

Linear plasma devices used to test plasma facing materials (PFMs) and components for fusion reactors are often suffering under the production of intrinsic impurities from the plasma source system. Most linear plasma devices use internal electrodes (hollow cathodes, reflex arc, or cascading arc), which typically are the source of the main impurities. The next-generation plasma generators use radio frequency (RF) plasma sources like helicons to avoid internal electrodes. However, high power operation of helicons has proven to result in impurity production due to the high rectified sheath voltages created. Depending on the plasma parameters and magnetic configuration, these impurities can be transported to the target and deposited there to unacceptable high levels. Here, in this contribution, the experimental results from Proto-MPEX are summarized, and the conclusions of the impurity source physics are given. Methods to reduce the impurity production, the impurity transport, and the net deposition on the target are presented. These methods to reduce the impurity production include Faraday screens to reduce the sheath voltage drop, high-Z refractory coatings to reduce the erosion yield, and wall conditioning methods. Methods to reduce the impurity transport include changes in the magnetic configuration as well as electron heating to change axial and radial temperature profiles. Preliminary results on the effectiveness of some of these methods are presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Enhancement of mid-/high-Z impurity transport by continuous Li-granule dropping in a stellarator plasma

An enhancement of core impurity transport is observed in high-density plasmas of the stellarator large helical device, heated by neutral beam injection, when continuous lithium (Li) granule dropping is performed. In the reported experiments, in which the tracer-encapsulated solid pellet is employed to inject trace amounts of titanium (Ti) and molybdenum (Mo) into the plasma core, confinement times for these impurities are seen to reduce significantly when Li dropping is applied, this reduction being more notable for Mo. To gain some initial insight into these observations, simulations are performed using the drift-kinetic transport code SFINCS for the Mo case. These simulations indicate that, while neoclassical transport prevails for the main plasma components (electrons, majority ions, and low-Z impurities), the classical contribution appears dominant for transporting Mo impurities. In summary, this work reports the first experimental observation of the degradation of mid-Z and high-Z impurity confinement induced by the continuous dropping of Li granules into a high-density stellarator plasma. In the case of the Mo impurity, simulations suggest that classical transport is the key mechanism underlying the enhanced impurity transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Simulation of plasma transport in MPS-LD linear plasma device by using BOUT++

Here a linear plasma device (LPD) module based on the 2D transport module under the BOUT++ framework is developed in this paper to simulate plasma transport in the LPD. The LPD module includes three parts, i.e. magnetic field calculation, simulation mesh generation and plasma transport setup. The magnetic field is calculated with the circular current loop using the location and current information of each coil. The mesh generation code can produce a simulation mesh for LPD by employing the magnetic field. The plasma transport model is based on the reduced Braginskii equations, which consist of the continuity equation, momentum equation, and energy equation. The fluid neutral model is applied for neutral particle simulation. Deuterium (D) and helium (He) ions and atoms can be simulated by the model. The first attempt to simulate plasma transport in the new LPD, called multiple plasma simulation linear plasma device (MPS-LD), is presented using the developed model. The D plasma transport in the MPS-LD is simulated and benchmarked against the two-point model, showing the validation of the BOUT++ simulation. The effects of radial transport on the heat load to the target are studied, which illustrates the significant impact of the radial diffusivities $D_⟂$ and $\mathcal{X}_{⟂i,e}$ on the plasma. Moreover, the He impurity injection process during the discharge is studied with emphasis on plasma–impurity interactions. The simulation results show that He injection can reduce the plasma energy load to the target significantly, and the efficiency depends on the He source density and injection velocity. The present work provides an alternative and flexible simulation tool for plasma transport in LPDs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1977

Oxygen impurity radiation from Tokamak-like plasmas

We have constructed a nonhydrodynamic coronal model for calculating radiation from impurity atoms in a heated plasma. Some recent developments in the calculation of dielectronic recombination rate coefficients and collisional excitation rate coefficients are included. The model is applied to oxygen impurity radiation during the first few milliseconds of a TFR Tokamak plasma discharge, and good agreement with experimental results is obtained. Estimates of total line and continuum radiation from the oxygen impurity are given. It is shown that impurity radiation represents a considerable energy loss.

Rogerson, J. E.↗
DOE OSTI2023

First feedback-controlled divertor detachment in W7-X: Experience from TDU operation and prospects for operation with actively cooled divertor

In the last experimental campaign (OP1.2b) of the stellarator Wendelstein 7-X (W7-X), boronisation as a mean for first wall conditioning was applied for the first time which led to strongly reduced impurity fluxes from plasma-facing components. Thermal detachment at the uncooled target plates of the test divertor unit (TDU) was reached at higher plasma densities and was accompanied by high recycling of neutrals at the target plate. A feedback control system was established in W7-X to actively control the gas injection (actuator) for plasma fuelling and impurity seeding [3] through the divertors. It allowed very successful stabilisation of the detached plasma condition state as well as mitigation of thermal overloads to some baffle tiles. Different routinely available diagnostic signals were used as input parameters (sensors). We describe the setup of the feedback control system, its performance and provide some example results with the main focus on the development of the control scheme which led to the detachment stabilisation over the entire longest (30 s) high-power discharge at W7-X so far. In view of the achieved very successful detachment stabilisation and the necessity to include simultaneous optimisation of the core performance in the future, the feedback control system is being upgraded for the upcoming campaign (OP2.1) in which the water cooled and hereby inherently steady-state capable divertor has been currently installed. The prospects and some experiment ideas for active detachment control are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Basic Physical Processes Involving Dust in Fusion Plasmas

This report presents the main outcomes of our research focused on understanding the behavior and effects of dust particles in fusion plasmas, particularly in the edge regions of tokamaks and stellarators. We conducted computer modeling studies of burst injections of carbon and tungsten dust particles in DIII-D like divertor plasmas. The studies investigated effects of transient influx of the low- and high-Z material plasma contaminants in form of dust on the edge plasma dynamics in a modern mid-size tokamak. We also performed theoretical and computational studies of the forces acting on non-spherical dust grains in magnetized and nonmagnetized plasmas. In addition, we cooperated with General Atomics slag management group on development of DIII-D dust injection experiments to measure trajectories of dust grains in the divertor plasma and compare them with DUSTT code predictions for validation of the dust modeling capabilities. We collaborated with JET experimentalists on evaluation of radiative losses induced by injection of mixed neon-deuterium ice pellets for disruption and run-away electron generation mitigation during thermal and current quench phases. We also cooperated with experimentalists at LHD fusion device (Japan) on modeling support of experiments on injection of boron granules in fusion plasma discharges to assess their dynamics and effectiveness for in situ dynamic wall conditioning.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Krash-ESL-Final report

During the period of the current Grant, the effort of the UCSD researchers was focused on: i) The application of the gyrokinetic code COGENT code to study the kinetic drift wave instability and its impact on blob dynamics, and ii) The modeling of edge plasma transport in collisional and semi-collisional regimes, and iii) Impurity transport in edge plasma and the incorporation of plasma-impurity collision operator (hereafter unlike particle collision operator) into COGENT. Over the years four UCSD graduate students (J. Angus, W. Lee, Y. Zhang, and A. Knyazev), funded under this Grant, contributed to the COGENT development/utilization and defended their PhD theses, which are based, partly, on their research related to the COGENT. Three of them (J. Angus, W. Lee, and Y. Zhang) have received “The Best MAE Graduate Student of the Year” awards. Whereas W. Lee after defending his PhD thesis returned back to his home country (South Korea), J. Angus, and Y. Zhang are continuing their research in plasma physics in the LLNL and LANL, whereas A. Knyazev currently is a postdoctoral fellow at Columbia University working on the theoretical aspects of advanced stellarators. The results obtained in our studies were disseminated in many papers published in first-class refereed scientific journals and presented on prestigious scientific meetings including the Annual APS DPP Meeting, Sherwood Fusion Theory Conference, IAEA Fusion Energy Conferences, International Conferences on Plasma-Surface Interactions in Controlled Fusion Devices, International Workshops on Edge Plasma Theory in Fusion Devices, etc. In what follows we highlight the major accomplishments of the UCSD Team in the main areas of our research.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Investigation of divertor detachment induced through neon seeding and density ramp on HL-3

A new self-consistent 1D scrape-off layer model has been recently developed in BOUT++ framework, named SD1D, which includes equations for various particle species (e.g. main plasma, neutrals and impurities) and couples open databases like ADAS and AMJUEL. It is able to quickly and effectively simulate divertor detachment experiments. In this work, a typical detachment experiment (shot #6270) on HL-3 with neon seeding is simulated using the SD1D code. It is found that the target electron temperature and the target ion saturation current in the simulations are consistent with experimental results measured by Langmuir probes on the target plate. The variation of D α radiation intensity in the divertor is qualitatively similar to the measured D α signal. Following the experimental validations, different upstream densities are set in the simulations to study the impurity distribution under different plasma density conditions. It is found that increasing upstream density can be helpful for the control of the neon radiation front (closer to the target). In this work we also compare two detachment regimes in simulations. Based on the same initial experimental parameters (shot #6270) on HL-3, a scan of upstream density and a scan of neon seeding rate are carried out respectively. It is found that the role of atomic and molecular processes is different in the two detachment regimes. The current density roll-over is ascribed to a drop in the divertor ion source, and the variation of D α radiation intensity via different excitation channels is associated with the relevant collisional reaction sources.

BOUT++↗
DOE OSTI2025

Active Divertor Heat Flux Control using Impurity Powder Dropper

Divertor plasma-facing components (PFCs) in a tokamak are typically designed to withstand average steady-state heat loads of about 5–10 MW/m 2 , a limit that applies to both solid and liquid lithium (LL) PFCs. Exceeding these design values can result in surface damage to tungsten PFCs or excessive lithium (Li) evaporation in liquid lithium divertor (LLD) PFCs. Since exceeding the divertor heat load limits has serious consequences, it is therefore prudent to develop a tool to reduce the divertor heat load and bring the heat load to within the design limit without affecting the plasma performance. Active low Z impurity injection such as Li has been suggested as a potential solution to mitigate excess heat flux as suggested previously, given that non-coronal radiation can be quite large ~ 20–30 MJ per mole of injected Li. Li is considered desirable for reducing the edge neutral recycling helping to improve plasma energy confinement. In this paper, we model the Impurity Power Dropper (IPD) to investigate its potential of divertor heat flux control. The IPD is typically located at the top of the tokamak device and uses a vertical drift tube of a few meters. In the 2 m drift tube case, the IPD powder is accelerated to ~ 6 m/sec before reaching the plasma with the upper divertor configuration, matching the condition for the in-board side pellet injection case. By modeling the IPD geometry we determined the IPD powder deposition profile, and thus the non-coronal radiation and ionization profiles in time as well. From the enhanced radiation power loss, it is therefore possible to reduce the divertor heat load using the divertor simulation code. In conclusion, the IPD divertor heat flux control can be tested in the facilities with IPD including ST-40, DIII-D, EAST, WEST and NSTX-U.

Active lithium injection↗
DOE OSTI2022

Ex-Situ Surface Characterization Studies and Boundary Plasma Diagnostic Development for DIII-D (Final Report)

This report details the accomplishments for award DE-SC0016318, which sponsored collaborative research activities between the University of Tennessee-Knoxville and the DIII-D experiment at General Atomics, led by the PI (Donovan), which officially began on August 1, 2016 and ended July 31, 2020. Though the official start date for the award was August 1, 2016, the PI had already initiated collaborative activities with DIII-D during 2015 supported by internal UTK start-up funds utilized by the PI. This prior work enabled UTK to have a substantial role in the June 2016 Metal Rings Campaign (MRC). The DOE funds then provided the opportunity to expand the UTK team with funding for students and a postdoc to perform ex-situ analysis on the wide array of samples exposed during the MRC and develop more sophisticated analysis tools and interpretive modeling techniques.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗