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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

The impact of non-local parallel electron transport on plasma-impurity reaction rates in tokamak scrape-off layer plasmas

Abstract Plasma-impurity reaction rates are a crucial part of modelling tokamak scrape-off layer (SOL) plasmas. To avoid calculating the full set of rates for the large number of important processes involved, a set of effective rates are typically derived which assume Maxwellian electrons. However, non-local parallel electron transport may result in non-Maxwellian electrons, particularly close to divertor targets. Here, the validity of using Maxwellian-averaged rates in this context is investigated by computing the full set of rate equations for a fixed plasma background from kinetic and fluid SOL simulations. We consider the effect of the electron distribution as well as the impact of the electron transport model on plasma profiles. Results are presented for lithium, beryllium, carbon, nitrogen, neon and argon. It is found that electron distributions with enhanced high-energy tails can result in significant modifications to the ionisation balance and radiative power loss rates from excitation, on the order of 50%–75% for the latter. Fluid electron models with Spitzer-Härm or flux-limited Spitzer-Härm thermal conductivity, combined with Maxwellian electrons for rate calculations, can increase or decrease this error, depending on the impurity species and plasma conditions. Based on these results, we also discuss some approaches to experimentally observing non-local electron transport in SOL plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Integrated modelling: Coupling of surface evolution and plasma-impurity transport

During the interaction of the scrape off layer (SOL) plasma with the first wall the evolution of both wall and plasma are tightly coupled: The erosion of the first wall leads to an impurity concentration in the plasma which affects the particle and power balance in the plasma. In turn the impurities, when leaving the plasma via transport, can form deposits and mixed materials, far away from their initial source location. These deposits can be eroded, allowing the impurities to stepwise migrate through the fusion device until they end up at a location where the plasma at the wall is cold enough and no further erosion occurs. To describe these processes an integrated model of surface evolution and plasma transport of impurities is needed. The WallDYN code achieves this required coupling of processes by parameterising the output of surface evolution- and plasma impurity-migration-codes by analytical models. For a given fixed background plasma it evolves the surface composition, derives impurity flux into and from the plasma and can from this derive the impurity densities in the plasma. This paper will show the importance of including this recycling of impurities at the wall in impurity migration modelling: The 13 CH 4 seeding experiment performed in the Wendelstein 7-X Stellerator is modelled using the recent extension of the WallDYN code to 3D plasma and wall geometries. A comparison with post mortem analysis of the 13 C deposition shows both qualitative and quantitative agreement with the WallDYN calculations.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The influence of surface impurities on photoelectric currents driven by intense soft x rays

In an x-ray driven cavity experiment, an intense flux of soft x rays on the emitting surface produces significant emission of photoelectrons having several kiloelectronvolts of kinetic energy. At the same time, rapid heating of the emitting surface occurs, resulting in the release of adsorbed surface impurities and subsequent formation of an impurity plasma. This numerical study explores a simple model for the photoelectric currents and the impurity plasma. In this work, attention is given to the effect of varying the composition of the impurity plasma. The presence of protons or hydrogen molecular ions leads to a substantially enhanced cavity current, while heavier plasma ions are seen to have a limited effect on the cavity current due to their lower mobility. Additionally, it is demonstrated that an additional peak in the current waveform can appear due to the impurity plasma. A correlation between the impurity plasma composition and the timing of this peak is elucidated.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Error-Field Penetration Thresholds in Ohmically Heated ITER and SPARC Plasmas

The critical n=1, n=2, and n=3 error-field amplitudes needed to trigger error-field penetration in ITER and SPARC, steady-state, ohmically heated plasmas are calculated using a standard asymptotic matching approach. The calculation incorporates plasma impurities, trapped particles, the bootstrap current, and neoclassical poloidal flow-damping. The energy confinement time is specified by the neo-Alcator scaling law in the low-density linear ohmic confinement (LOC) regime, and by the ITER-89P L-mode scaling law in the high-density saturated ohmic confinement regime (SOC) regime. The response of the plasma in the inner region is calculated using a linearized version of the four-field model. At the normal operating electron number density, diamagnetic levels of rotation are found to be sufficient to protect ITER and SPARC ohmically heated plasmas from m=2/n=1 error-field penetration. On other hand, SPARC, and especially ITER, ohmically heated plasmas may be vulnerable to n>1 error-field penetration. ITER and SPARC ohmically heated plasmas are also slightly more susceptible to error-field penetration when the electron fluid at the rational surface rotates in the ion diamagnetic direction, rather than the electron diamagnetic direction. At electron number densities that are sufficiently low that the plasma lies in the LOC confinement regime, the error-field penetration threshold increases with increasing density. However, as soon as the electron number density becomes large enough that the plasma enters the SOC regime, the increase in the error-field penetration threshold with increasing density levels off.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Time-dependent modeling of dust outburst into tokamak divertor plasma

The first self-consistent simulation studies of the impact of transient injection of tungsten dust on ITER-like edge plasma are presented. The simulations are performed using DUSTT and UEDGE codes coupled in a fully time-dependent manner. The evolution of the divertor and upstream plasma conditions, following the injection of tungsten dust of various quantities and sizes at the outer strike point, is investigated. The estimates of the core plasma impurity fraction and the core impurity screening factor associated with the dust injection are also obtained. Here, the simulations show that the bursts of even modest amounts of tungsten dust injected into the divertor plasmas can result in large perturbations of the divertor power load and unacceptable levels of the core plasma contamination with tungsten impurities.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Modification of ion-temperature-gradient turbulence by impurities in stellarator plasmas

Recent nonlinear gyrokinetic simulations have shown that impurities can strongly modify the turbulent heat flux in stellarator plasmas. Here, the ion-temperature-gradient (ITG) dispersion relation in a plasma containing impurities is analytically solved in certain limits and an expression for the modification of the ITG growth rate by impurities is derived. The analytical expression is the sum of three terms corresponding to three different physical causes (impurity density gradient, impurity temperature gradient and dilution) of the change in the growth rate. The scalings predicted analytically for the modification of the growth rate are shown to be reproduced by linear gyrokinetic simulations. The conditions for reduction or increase of the ITG growth by impurities are also correctly predicted by the analytical solution to the dispersion relation. Finally, a remarkable correlation is found between the analytical expression for the modification of the growth rate and the modification of the turbulent heat flux obtained from nonlinear gyrokinetic simulations.

Calvo, Iván [Research Centre for Energy, Environme↗
DOE OSTI2020

Interpretative transport modeling of the WEST boundary plasma: main plasma and light impurities

Understanding impurity transport in tokamak plasmas is crucial to control radiative losses and material migration in future magnetic fusion reactors. In this work we deploy the SolEdge2D-EIRENE code to model the boundary plasma in a WEST discharge, satisfactorily reproducing measurements of both upstream and divertor plasma conditions. Additionally, the spatial distribution of oxygen, studied here as a representative light impurity, is compared to vacuum ultraviolet spectroscopy measurements acquired with an oscillating line of sight. The simulation captures a key feature of the experiment, namely a factor of sime2 higher oxygen brightness in the inner divertor region compared to the outer one. This spatial asymmetry in oxygen concentration is interpreted by analyzing the balance of friction forces and thermal gradient forces that the light impurity exchanges with the main plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Initial design concepts for solid boron injection in ITER

As part of ITER’s consideration to change its first wall material from beryllium to tungsten, the ITER Organization has proposed studying the feasibility of real-time solid boron injection (SBI) into the plasma to coat the walls and divertor to supplement glow discharge boronization (GDB). Boron deposits getter oxygen and reduce sputtering of tungsten from plasma facing components (PFCs). Particularly in areas with significant plasma wall interactions, boron coatings are expected to be short-lived under high performance plasma conditions. The proposed SBI system aims to maintain boron layers in these areas to avoid excessive radiation from tungsten in the plasma as a risk mitigation to ensure ITER will be able to reach and sustain Q = 10 conditions. The system will be used sparingly, as redeposition of boron can lead to significant tritium retention, which must be minimized in ITER to comply with nuclear safety concerns. SBI is proposed to limit and precisely control the amount of boron injected in real-time during plasma operation. Here, some of the design requirements and initial concepts for an SBI system in ITER are presented based on previous results carried out with SBI systems on a number of tokamaks and stellarators around the world. Previous results using SBI systems installed by PPPL have injected boron particles from 5 µm–2 mm diameter at calibrated rates of 2–200 mg/s in real-time during plasma operation on AUG, DIII-D, EAST, KSTAR, LHD, TFTR, WEST, and W7-X, leading to improved wall conditions with reduced plasma impurity concentrations and radiated power and improved plasma performance. The boron is ionized in the plasma edge and then deposited on plasma-wetted surfaces. On AUG, EAST and WEST reduced tungsten sputtering sources were observed following several discharges with SBI. Extrapolation of these SBI results are presented to estimate the amount of boron needed for wall conditioning in ITER. Real-time SBI control requirements and plasma operation scenarios for ITER are also described.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Gyrokinetic prediction of core tungsten peaking in a WEST plasma with nitrogen impurities

Tungsten peaking is predicted in the core of a WEST plasma with total-f gyrokinetic simulations, including both collisional and turbulent transport. This prediction is validated with a synthetic diagnostic of the bolometry. Although nitrogen impurities are shown to reduce the neoclassical peaking of tungsten on-axis, the overall tungsten peaking increases when nitrogen impurities are present, as they reduce the turbulence screening off-axis. This finding is important for the plasma current ramp-up phase of ITER, where light impurities seeding will be desirable to achieve low temperatures at the plasma-facing components and reduce tungsten sputtering. It provides further argument for applying early ECRH heating to maintain margins on the core power balance. The neoclassical peaking factor is cross-verified between XGC and FACIT. The heat flux at separatrix and the heat load width are modeled by XGC and compared to WEST data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

A 1D Lyman-alpha Profile Camera for Plasma Edge Neutral Studies on the DIII-D Tokamak

A one dimensional, absolutely calibrated pinhole camera system was installed on the DIII-D tokamak to measure edge Lyman-alpha (Ly-a) emission from hydrogenic isotopes which can be used to infer neutral density and ionization rate proles. The system is composed of two cameras, each providing a toroidal fan of twenty lines of sight, viewing the plasma edge on the inboard and outboard side of DIII-D. The cameras' views lie in a horizontal plane 77 cm below the midplane. At its tangency radius, each channel provides a radial resolution of approximately 2 cm full width at half maximum (FWHM) with a total coverage of 22 cm. Each camera consists of a rectangular pinhole, Ly-a reflective mirror, narrow-band Ly-a transmission fiter, and a 20 channel AXUV photodetector. The combined mirror and transmission fiter have a FWHM of 5 nm, centered near the Ly-a wavelength of 121.6 nm and is capable of rejecting significant, parasitic carbon-III (C-III) emission from intrinsic plasma impurities. To provide a high spatial resolution measurement in a compact footprint, the camera utilizes advanced engineering and manufacturing techniques including 3D printing, high stability mirror mounts, and a novel alignment procedure. Absolutely calibrated, spatially resolved Ly-a brightness measurements utilize a bright, isolated line with low parasitic surface reflections and enable quantitative comparison to modeling to study divertor neutral leakage, main chamber fueling and radial particle transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Implementation of AI/DEEP learning disruption predictor into a plasma control system

Abstract This paper reports on advances in the state‐of‐the‐art deep learning disruption prediction models based on the Fusion Recurrent Neural Network (FRNN) originally introduced in a 2019 NATURE publication [ https://doi.org/10.1038/s41586‐019‐1116‐4 ]. In particular, the predictor now features not only the “disruption score,” as an indicator of the probability of an imminent disruption, but also a “sensitivity score” in real time to indicate the underlying reasons for the imminent disruption. This adds valuable physics interpretability for the deep learning model and can provide helpful guidance for control actuators now implemented into a modern plasma control system (PCS). The advance is a significant step forward in moving from modern deep learning disruption prediction to real‐time control and brings novel AI‐enabled capabilities relevant for application to the future burning plasma ITER system. Our analyses use large amounts of data from JET and DIII‐D vetted in the earlier NATURE publication. In addition to “when” a shot is predicted to disrupt, this paper addresses reasons “why” by carrying out sensitivity studies. FRNN is accordingly extended to use more channels of information, including measured DIII‐D signals such as (i) the “n1rms” signal that is correlated with the n = 1 modes with finite frequency, including neoclassical tearing mode and sawtooth dynamics; (ii) the bolometer data indicative of plasma impurity control; and (iii) “q‐min”—the minimum value of the safety factor relevant to the key physics of kink modes. The additional channels and interpretability features expand the ability of the deep learning FRNN software to provide information about disruption subcategories as well as more precise and direct guidance for the actuators in a PCS.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

A 1D Lyman-alpha profile camera for plasma edge neutral studies on the DIII-D tokamak

A one dimensional, absolutely calibrated pinhole camera system was installed on the DIII-D tokamak to measure edge Lyman-alpha (Ly-α) emission from hydrogen isotopes, which can be used to infer neutral density and ionization rate profiles. The system is composed of two cameras, each providing a toroidal fan of 20 lines of sight, viewing the plasma edge on the inboard and outboard side of DIII-D. The cameras’ views lie in a horizontal plane 77 cm below the midplane. At its tangency radius, each channel provides a radial resolution of ~2 cm full width at half maximum (FWHM) with a total coverage of 22 cm. Each camera consists of a rectangular pinhole, Ly-α reflective mirror, narrow-band Ly-α transmission filter, and a 20 channel AXUV photodetector. The combined mirror and transmission filter have a FWHM of 5 nm, centered near the Ly-α wavelength of 121.6nm and is capable of rejecting significant, parasitic carbon-III (C-III) emission from intrinsic plasma impurities. To provide a high spatial resolution measurement in a compact footprint, the camera utilizes advanced engineering and manufacturing techniques including 3D printing, high stability mirror mounts, and a novel alignment procedure. Absolutely calibrated, spatially resolved Ly-α brightness measurements utilize a bright, isolated line with low parasitic surface reflections and enable quantitative comparison to modeling to study divertor neutral leakage, main chamber fueling, and radial particle transport.

Ionization Rate↗
DOE OSTI2023

Enhanced plasma current spike formation due to onset of 1/1 kink-tearing reconnection during a massive gas injection process

The formation of the plasma current spike at the end of the thermal quench (TQ) phase is studied systematically, which is found to strongly correlate with the onset of the 1/1 kink-tearing reconnection only in the simulation results presented here. Here, the magnetohydrodynamic activity on the q = 1 surface plays a critical role in the spike formation and the disruption process. In particular, when the safety factor in the magnetic axis q 0 exceeds 1, the plasma major disruption transits into successive minor disruptions, and the start of the TQ phase is delayed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Dependence of high- Z redeposition on the field-to-surface pitch angle and other sheath parameters in tokamaks

Accurately predicting redeposition is vital for high-Z plasma-facing component (PFC) survivability in magnetic confinement fusion. In this study, we categorize high-Z redeposition into three mechanisms: geometric-driven (prompt), sheath-driven (local), and scrape-off-layer-driven (far) redeposition. To investigate these mechanisms, we employ Monte Carlo transport codes to simulate azimuthally symmetric tungsten source erosion and redeposition in a tokamak. By iteratively analyzing critical parameters, we evaluate redeposition scaling for each mechanism. Specifically, we investigate the impact of magnetic-field-to-PFC pitch angle assumptions on PFC losses into the scrape-off layer. Our findings reveal significant pitch angle sensitivity due to an asymmetric prompt vs local redeposition trade-off. These results enhance our understanding of redeposition phenomena in fusion plasma environments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗