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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 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Power Balance Estimation in Long Duration Discharges on QUEST

Fully non-inductive plasma start-up was successfully achieved by using a well-controlled microwave source on the spherical tokamak, QUEST. Non-inductive plasmas were maintained for approximately 3–5 min, during which time power balance estimates could be achieved by monitoring wall and cooling-water temperatures. Approximately 70%–90% of the injected power could be accounted for by calorimetric measurements and approximately half of the injected power was found to be deposited on the vessel wall, which is slightly dependent on the magnetic configuration. Lastly, the power distribution to water-cooled limiters, which are expected to be exposed to local heat loads, depends significantly on the magnetic configuration, however some of the deposited power is due to energetic electrons, which have large poloidal orbits and are likely to be deposited on the plasma facing components.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Drift effects on W7-X divertor heat and particle fluxes

Classical particle drifts are known to have substantial impacts on fluxes of particles and heat through the edge plasmas in both tokamaks and stellarators. Here we present results from the first dedicated investigation of drift effects in the W7-X stellarator. By comparing similar plasma discharges conducted with a forward- and reverse-directed magnetic field, the impacts of drifts could be isolated through the observation of up-down asymmetries in flux profiles on the divertor targets. In low-density plasmas, the radial locations of the strike lines (i.e. peaks in the target heat flux profiles) exhibited discrepancies of up to 3 cm that reversed upon magnetic field reversal. In addition, asymmetric heat loads were observed in regions of the target that are shadowed by other targets from parallel flux from the core plasma. A comparison of these asymmetric features with the footprints of key topological regions of the edge magnetic field on the divertor suggests that the main driver of the asymmetries at low density is poloidal E × B drift due to radial electric fields in the scrape-off layer and private flux region. In higher-density plasmas, upper and lower targets collected non-ambipolar currents with opposite signs that also inverted upon field reversal. Overall, in these experiments, almost all up-down asymmetry could be attributed to the field reversal and, therefore, field-dependent drifts.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

The achievement of the T e,div feedback control by CD 4 seeding on EAST

A multi-function divertor feedback control system has been built on Experimental Advanced Superconducting Tokamak (EAST) to treat the divertor heat load issue. With the real-time data of the Langmuir probes and the impurity seeding, the divertor electron temperature (T e,div ) is well controlled to achieve the partial detachment phase. The first trial of the CD 4 seeding for the T e,div reduction has been achieved on EAST long-pulse discharge. In the seeding phase, the T e,div was maintained close to 5 eV, and the surface temperature of the target plate (T surface,div ) had a reduction of about 150° C. The plasma stored energy had a reduction in the control phase, so it is necessary to find a way to keep the good plasma confinement in the next step. The CD 4 injection also mitigated the low hybrid wave coupling rate in some degree. The big volume of the CD 4 injection lifted the Greenwald density fraction from ~0.4 to ~0.7, which made the SOL into high recycling state. As a result, most of the injected carbon particles were in the high ionized state, and with the lower of the T e,div , the tungsten line emission was suppressed obviously.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Ion temperature anisotropy in the tokamak scrape-off layer

Understanding tokamak exhaust-power heat loads on divertor plates depends critically on having a realistic model of the scrape-off layer (SOL) plasma. The Braginskii fluid model is often solved to understand the SOL plasma behavior. This model is based on the collisional limit for transport along the magnetic field B → . The ions and electron gyrofrequencies are assumed to be much larger than the Coulomb collision frequencies, which are nonetheless, sufficiently large to yield common parallel and perpendicular temperatures for each species, i.e. the temperatures are assumed to be isotropic. Additionally, in certain circumstances such as encountered for the tokamak H-mode, the ion temperature can be quite anisotropic. In this work, the anisotropy effects are implemented in the two-dimensional (2D) transport code UEDGE. Various geometries (1D slab, 2D slab and a toroidal tokamak geometry) are used to study the 2D structure of ion temperature anisotropy and its effects on plasma transport in detail. Results show that the effects of ion temperature anisotropy on the plasma parallel transport are substantial near the magnetic X-point, which leads to different steady state density profiles in the divertor regions. The extra mirror force introduced by ion temperature anisotropy can be one of the main forces contributing to the plasma flow in the SOL.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Physics of runaway electrons with shattered pellet injection at JET

Runaway electrons (REs) created during tokamak disruptions pose a threat to the reliable operation of future larger machines. Experiments using shattered pellet injection (SPI) have been carried out at the JET tokamak to investigate ways to prevent their generation or suppress them if avoidance is not sufficient. Avoidance is possible if the SPI contains a sufficiently low fraction of high-Z material, or if it is fired early in advance of a disruption prone to runaway generation. These results are consistent with previous similar findings obtained with Massive Gas Injection. Suppression of an already accelerated beam is not efficient using High-Z material, but deuterium leads to harmless terminations without heat loads. This effect is due to the combination of a large magnetohydrodynamic instability scattering REs on a large area and the absence of runaway regeneration during the subsequent current collapse thanks to the flushing of high-Z impurities from the runaway companion plasma. This effect also works in situations where the runaway beam moves upwards and undergoes scraping-off on the wall.

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↗
DOE OSTI2025

Runaway electron-induced plasma facing component damagein tokamaks

This Roadmap article addresses the critical and multifaceted challenge of plasma facing component (PFC) damage caused by runaway electrons (REs) in tokamaks, a phenomenon that poses a significant threat to the viability and longevity of future fusion reactors such as ITER and DEMO. The dramatically increased RE production expected in future high-current tokamaks makes it very difficult to avoid or mitigate REs in such devices when a plasma discharge terminates abnormally. Preventing damage from the intense localised heat loads they can cause requires a holistic approach that considers plasma, REs and PFC damage. Despite decades of progress in understanding the physics of REs and the thermomechanical response of PFCs separately, their complex interplay remains poorly understood. This document aims to initiate a coordinated, interdisciplinary approach to bridge this gap by reviewing experimental evidence, advancing diagnostic capabilities, and improving modelling tools across different scales, dimensionalities, and fidelities. Key topics include RE beam formation and transport, damage mechanisms in both brittle and metallic PFCs, and observed effects in major facilities such as JET, DIII-D, WEST and EAST. The Roadmap emphasises the urgency of predictive, high-fidelity modelling validated against well-diagnosed controlled experiments, particularly in the light of recent changes in ITER’s wall material strategy and the growing importance of private sector fusion initiatives. Each section of the Roadmap article is written to provide a concise overview of one area of this multidisciplinary subject, with an assessment of the status, a look at current and future challenges, and a brief summary. The ultimate goal of this initiative is to guide future mitigation strategies and design resilient components that can withstand the intense localised loads imposed by REs, thus ensuring the safe and sustainable operation of the next generation of fusion power plants.

Ratynskaia, Svetlana [KTH Royal Inst. of Technolog↗
DOE OSTI2017

Next generation HOM-damping

Research and development for superconducting radio-frequency cavities has made enormous progress over the last decades from the understanding of theoretical limitations to the industrial mass fabrication of cavities for large-scale particle accelerators. Key technologies remain hot topics due to continuously growing demands on cavity performance, particularly when in pursuit of high quality beams at higher beam currents or higher luminosities than currently achievable. This relates to Higher Order Mode (HOM) damping requirements. Meeting the desired beam properties implies avoiding coupled multi-bunch or beam break-up instabilities depending on the machine and beam parameters that will set the acceptable cavity impedance thresholds. The use of cavity HOM-dampers is crucial to absorb the wakefields, comprised by all beam-induced cavity Eigenmodes, to beam-dynamically safe levels and to reduce the heat load at cryogenic temperature. Cavity damping concepts may vary, but are principally based on coaxial and waveguide couplers as well as beam line absorbers or any combination. Next generation Energy Recovery Linacs and circular colliders call for cavities with strong HOM-damping that can exceed the state-of-the-art, while the operating mode efficiency shall not be significantly compromised concurrently. This imposes major challenges given the rather limited damping concepts. A detailed survey of established cavities is provided scrutinizing the achieved damping performance, shortcomings, and potential improvements. The scaling of the highest passband mode impedances is numerically evaluated in dependence on the number of cells for a single-cell up to a nine-cell cavity, which reveals the increased probability of trapped modes. This is followed by simulations for single-cell and five-cell cavities, which incorporate multiple damping schemes to assess the most efficient concepts. The usage and viability of on-cell dampers is elucidated for the single-cell cavity since it can push the envelope towards quasi HOM-free operation suited for next generation storage and collider rings. Geometrical end-cell shape alterations for the five-cell cavity with already efficient mode damping are discussed as a possibility to further lower specific high impedance modes. Lastly, the findings are eventually put into relation with demanding impedance instability thresholds in future collider rings.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

Ultra-low magnetization and hysteresis loss in APC Nb 3 Sn superconductors

For the accelerator magnets of the next hadron collider, reducing superconductor persistent-current magnetization is not only important for achieving the desired field quality, but also crucial for its sustainability because the magnetization loss is the major heat load to the magnet cold mass. For conventional Nb 3 Sn conductors this requires reduction of effective subelement size (D eff ). For the restacked-rod-process (RRP ® ) conductors a physical subelement size (D sub ) as small as 35 µm (corresponding to a D eff close to 45 µm) can be reached, but at a significant price in J c . Another way to reduce the magnetization is by introducing artificial pinning centers (APC) using the internal oxidation approach. APC conductors outperform conventional Nb 3 Sn wires in two aspects: 1) higher J c at high fields, and 2) much lower J c and magnetization at low fields (e.g. below 5 T). Here, in this work we explored the fabricability of APC wires with small D sub . A 180-stack APC wire was produced and drawn to 0.7- and 0.5 mm diameters with good quality, with D sub s of 34 and 24 µm (D eff s of 36 and 25 µm), respectively. For the 34 µm-D sub wire, its non-Cu J c is higher than that of an RRP ® wire used for the High-Luminosity Large Hadron Collider (HL-LHC) project above 13 T (e.g. 36% higher at 4.2 K, 18 T), while its non-Cu magnetization at 1 T, ΔM(1 T), is only 29% of the RRP ® wire. Its non-Cu hysteresis loss for a cycle between 1 and 14 T, Q h (1–14 T), is 37% of the RRP ® wire. For the 24 µm-D sub wire, its non-Cu J c surpasses the HL-LHC RRP ® wire above 17.5 T, while its ΔM(1 T) and Q h (1–14 T) are only 17% and 23% of the RRP® wire, respectively. Its non-Cu Q h (±3 T) even meets the specification of the International Thermonuclear Experimental Reactor project.

Xu, X [Fermi National Accelerator Laboratory (FNAL↗
DOE OSTI2019

Radiative edge cooling experiments in Wendelstein 7-X start-up limiter campaign

We report impurity seeding experiments during the start-up limiter campaign of Wendelstein 7-X provide first evidence for a localization effect of the 3D magnetic edge structure on the seeded impurities and their radiation distribution and cooling effect. Moreover, species dependencies have been seen. Nitrogen was observed to cool the entire edge plasma, with a stronger electron temperature reduction measured at the downstream position at the limiter. The radiation was limited at the periphery of the confined region. Both the temperature reduction and the radiation enhancement were directly correlated to the injection of the coolant gas. Mitigation of the limiter heat loads was also measured. Neon was observed to affect also the confined plasma with a long-lasting radiation and a cooling of the entire plasma.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Assessment of equilibrium field coil misalignments on the divertor footprints in NSTX-U

The presence of error fields in a tokamak device is inevitable, so it is of fundamental importance to understand their impact on the operation of the device and therefore up to what level they can be tolerated. In this paper a prediction of the impact on the magnetic footprints on the divertor plates due to the misalignment of NSTX-U equilibrium coils is presented. Resistive MHD solutions are used to predict the magnetic field perturbations experienced by the field lines due to the presence of error fields. A linear relation between the magnitude of the misalignment and the area of the magnetic footprints is found, as well as an inverse proportionality between the size of a magnetic footprint and the plasma penetration of the field lines composing the footprint. This study is intended as a starting point in the process of investigating divertor footprints associated with the plasma response to intrinsic error fields with resistive MHD codes. A basic interpretation of the results suggests that large footprints resulting from equilibrium coils misalignment may be desirable, to some extent, to decrease the peak heat load on the divertor components. Lastly, a more complete interpretation of the results presented here will involve a number of complex physics issues, for example pedestal transport and stability along with the role of various scrape-off layer and divertor processes, as well as experimental data to confirm the predictions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

An improved understanding of the roles of atomic processes and power balance in divertor target ion current loss during detachment

The process of divertor detachment, whereby heat and particle fluxes to divertor surfaces are strongly diminished, is required to reduce heat loading and erosion in a magnetic fusion reactor to acceptable levels. In this paper, the physics leading to the decrease of the total divertor ion current ($I_t$), or ‘roll-over’, is experimentally explored on the TCV tokamak through characterization of the location, magnitude and role of the various divertor ion sinks and sources including a complete analysis of particle and power balance. These first measurements of the profiles of divertor ionisation and hydrogenic radiation along the divertor leg are enabled through novel spectroscopic techniques. Over a range in TCV plasma conditions (plasma current and electron density, with/ without impurity-seeding) the $I_t$ roll-over is ascribed to a drop in the divertor ion source; recombination remains small or negligible farther into the detachment process. Further, the ion source reduction is driven by both a reduction in the power available for ionization, $P_{\text{recl}}$, and concurrent increase in the energy required per ionisation, $E_{\text{ion}}$: this effect of power available on the ionization source is often described as ‘power starvation’ (or ‘power limitation’). The detachment threshold is found experimentally (in agreement with analytic model predictions) to be ~$P_{\text{recl}}/I_tE_{\text{ion}}$ ~ 2, corresponding to a target electron temperature, $T_t ~ E_{\text{ion}}/γ$ where $\gamma$ is the sheath transmission coefficient. The target pressure reduction, required to reduce the target ion current, is driven both by volumetric momentum loss as well as upstream pressure loss. The measured evolution through detachment of the divertor profile of various ion sources/ sinks as well as power losses are quantitatively reproduced through full 2D SOLPS modelling through the detachment process as the upstream density is varied.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

A new mechanism for filament disconnection at the X-point: poloidal shear in radial E × B velocity

Plasma filaments generated by turbulence in the scrape-off layer are sometimes observed in experiments and simulations to disconnect from the target plate in the vicinity of the X-point, resulting in a ‘quiescent zone’ with reduced fluctuations, possibly affecting the fluctuating part of the divertor heat loads. This phenomena is usually explained by the flux tubes squeezing, induced by flux expansion, down to scales where collisions dominate. In this work, we consider an additional mechanism spontaneously arising at the X-point: the poloidal shear in radial E × B velocity. Through a simple model, tested against 3D global turbulence simulations, herein we quantify the effectiveness of this disconnection mechanism and identify ways to control it.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Divertor detachment in the pre-fusion power operation phase in ITER during application of resonant magnetic perturbations

Detachment of the divertor plasma during application of resonant magnetic perturbation (RMP) fields is evaluated for hydrogen H-mode plasma during the first pre-fusion power operation (PFPO-1) phase in ITER by 3D plasma boundary modelling with EMC3–EIRENE. Plasma response effects from a linearized, resistive, single fluid MHD model are discussed, which includes partial screening of the externally applied field—but also field amplification near the separatrix. Furthermore, this field amplification is found to play a pivotal role for the magnetic footprint on the divertor targets, but is sensitive to model parameters. Extensions of the footprint beyond the straight portions of the ITER vertical divertor targets, optimized for high stationary heat flux handling, may be possible depending on the level of toroidal rotation in the plasma. Exhaust from the bulk plasma is guided by the helical corrugations (lobes) of the perturbed separatrix, and this results in an upstream heat flux that is distributed over these lobes with lower peak values than in the typical radial heat flux profiles seen in the absence of magnetic perturbations. As a consequence, an earlier onset (with respect to the upstream density) of detachment is found in the traditional strike zone when RMPs are applied, but secondary, non-axisymmetric strike locations appear—and those remain attached at temperatures above 10 eV. Neon seeding can mitigate these non-axisymmetric heat loads, but this becomes less efficient for large magnetic footprints.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Disruptive neoclassical tearing mode seeding in DIII-D with implications for ITER

New studies identify the critical parameters and physics governing disruptive neoclassical tearing mode (NTM) onset. An m/n = 2/1 mode in DIII-D that begins to grow robustly after a seeding event (edge localized mode ELM or sawtooth precursor and crash) causes the mode rotation to drop close to the plasma's Er = 0 rest frame; this condition opens the stabilizing ion-polarization current 'gate' and destabilizes an otherwise marginally stable NTM. Our new experimental and theoretical insights and novel toroidal theory-based modeling are benchmarked and scalable to ITER and other future experiments. Here, the nominal ITER rotation at q = 2 is found to be stabilizing ('gate closed') except for MHD-induced transients that could 'open the gate'. Extrapolating from the DIII-D ITER baseline scenario (IBS) discharges, MHD transients are much more likely to destabilize problematic robustly growing 2/1 NTMs in ITER; this makes predictions of seeding and control of both ELMs and sawteeth imperative for more than just minimizing divertor pulsed-heat loading.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Tungsten–carbon surface evolution and erosion modeling for a small angle slot divertor in DIII-D

Abstract We modeled tungsten–carbon mixed surface evolution, sputtering erosion, and transport for the tungsten coated region of a small angle slot (SAS) divertor design for the DIII-D tokamak. This divertor concept aims to achieve a closed slot dissipative plasma to minimize heat load and surface erosion, and to study high- Z material performance. Our advanced simulations use coupled ITMC-DYN material mixing/response and 3D full kinetic REDEP/WBC erosion/redeposition code packages, with divertor plasma solution from the SOLPS-ITER package with 4 MW power input. The SAS design geometry and resulting in-slot plasma parameters cause significant differences in predicted sputter/transport from a conventional divertor. For 2% C/D incident plasma ratio, an equilibrium mixed C/W surface is attained at ∼30 s of discharge, from wall sputtered carbon transported to the 10 cm long tungsten divertor region. Tungsten remains exposed to the plasma, but the evolved surface composition varies with different C/D ratios. Tungsten is primarily sputtered from the mixed surface by impinging carbon ions in the +1 to +4 charge states, with some self-sputtering. Redeposition of sputtered tungsten to the divertor is significant, ∼80% along the higher plasma temperature attached plasma SAS entrance region, but this is less than the typically near-unity values for a conventional divertor. Plasma-incident carbon is highly backscattered (∼50%) from the mixed surface, with little redeposition (<10%); this helps maintain tungsten in the surface sputter zone. Carbon is mainly sputtered from the mixed surface by D + ions, also with low redeposition (∼10%–30%). Finally, the modeling shows non-zero but low sputtered tungsten current from the divertor to the core plasma direction. These results appear favorable for effective testing of a tungsten-containing SAS divertor in DIII-D, and extrapolation of mixed-material evolution/response findings to the analogous low- Z /high- Z , Be/W, ITER plasma facing system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Impact of Soret effect on hydrogen and helium retention in PFC tungsten under ELM-like conditions

Abstract In our previous work, we have demonstrated using nonequilibrium molecular-dynamics simulations that the fluxes of helium and self-interstitial atoms in the presence of a thermal gradient in tungsten are directed opposite to the heat flux, indicating that species transport is governed by a Soret effect, namely, thermal-gradient-driven diffusion, characterized by a negative heat of transport that drives species transport uphill, i.e. from the cooler to the hot regions of the tungsten sample. In this work, the findings of our thermal and species transport analysis have been implemented in our cluster-dynamics code, Xolotl, which has been used to compute temperature and species profiles over spatiotemporal scales representative of plasma-facing component (PFC) tungsten under typical reactor operating conditions, including extreme heat loads at the plasma-facing surface characteristic of plasma instabilities that induce edge localized modes (ELMs). We demonstrate that the steady-state species profiles, when properly accounting for the Soret effect, vary significantly from those where temperature-gradient-driven transport is not accounted for and discuss the implications of such a Soret effect on the response to plasma exposure of plasma-facing tungsten. Although our cluster-dynamics simulations do not yet include self-clustering of helium or hydrogen blister formation, our simulation results show that the Soret effect substantially reduces helium and hydrogenic species retention inside PFC tungsten.

36 MATERIALS SCIENCE↗
DOE OSTI2023

The JET hybrid scenario in Deuterium, Tritium and Deuterium-Tritium

The JET hybrid scenario has been developed from low plasma current carbon wall discharges to the record-breaking Deuterium-Tritium plasmas obtained in 2021 with the ITER-like Be/W wall. The development started in pure Deuterium with refinement of the plasma current, and toroidal magnetic field choices and succeeded in solving the heat load challenges arising from 37 MW of injected power in the ITER like wall environment, keeping the radiation in the edge and core controlled, avoiding MHD instabilities and reaching high neutron rates. The Deuterium hybrid plasmas have been re-run in Tritium and methods have been found to keep the radiation controlled but not at high fusion performance probably due to time constraints. For the first time this scenario has been run in Deuterium-Tritium (50:50). These plasmas were re-optimised to have a radiation-stable H-mode entry phase, good impurity control through edge T i gradient screening and optimised performance with fusion power exceeding 10 MW for longer than three alpha particle slow down times, 8.3 MW averaged over 5 s and fusion energy of 45.8 MJ.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗