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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Development of an Integrated Multidiagnostic to Assess the High-Z Impurity Fluxes in the Metallic Environment of WEST Using IMAS

WEST is an actively cooled, long-pulse tokamak with nearly all plasma-facing components (PFC) made of tungsten. One of the aims of WEST is to study plasma operations with tungsten PFCs in preparation for long-pulse operations on high-Z divertor devices, such as ITER. For long-pulse operation, the high-Z impurity content and transport to the core plasma are critical concerns that require further measurement and interpretation to improve plasma performance and PFC durability. This work details the impurity influxes in WEST during a series of discharges in which the lower hybrid (LH) injected power was incrementally increased. An analysis has been performed of measurements collected from an array of edge diagnostics. Visible spectroscopy was utilized to measure the spectral radiances generated by fuel particles (D) and impurities (W, O, and C) at the divertor and at the antennas with a newly developed spectral peak-fitting tool used to analyze the data in WEST. The scrape-off layer (SOL) plasma conditions (n e and T e ) measured at the divertor target with flush Langmuir probes and near the outer mid-plane (OMP) using reciprocating Langmuir probes (RCPs) are used to evaluate the number of ionizations per photons (S/XB) coefficients required to estimate the impurity fluxes obtained with the collisional-radiative model ColRadPy. Here, the array of edge diagnostics discussed in this work, coupled with SOL plasma modeling tools, represents a multidiagnostic interpretative modeling workflow that will continue to be applied to upcoming experimental campaigns on the WEST experiment to assess high-Z impurity transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Effect of lithium coating on long pulse high performance plasma discharges in EAST

Control of impurities, fuel recycling and hydrogen content by lithium evaporative coatings and real-time lithium powder injection (LPI) in EAST are studied for high performance H-mode discharges of up to ~100 s. The results show that the lithium evaporative coatings significantly reduced both the low-Z impurity carbon and high-Z impurity tungsten as well as molybdenum concentration in the plasmas, and the impurities concentration significantly reduced with the accumulated lithium coatings and maintained well afterwards. Specifically the high-Z tungsten core impurity concentration was maintained between 3 ppm–15 ppm during the 101 s H-mode discharge, which is acceptable for the long pulse operation. In addition, real-time wall conditioning via LPI successfully reduced the core high-Z metal impurities by 50% during ~35 s long pulse H-mode discharge, exhibiting strong compatibility between real-time LPI with long pulse discharges. In addition, evaporative lithium coatings demonstrated fuel recycling control, with fuel recycling obviously reducing with lithium coatings and maintaining well afterwards. Also, the lithium evaporative coatings reduced the hydrogen minority species content, represented by the density ratio H/(H + D), from ~50% down to ~5% with accumulated lithium coatings; low hydrogen fraction improved the ICRF minority heating efficiency. Finally the real-time LPI reduced the recycling coefficient R global from 0.95 to 0.82. With these wall conditionings help, the plasma density controlled well during the 101 s long pulse H-mode discharge. Furthermore, these results provide valuable references on impurities, fuel recycling and hydrogen content control for future longer pulse high performance H-mode operation (≥400 s) in EAST and future fusion devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Fuel recycling and impurity characteristics in long-pulse H-mode plasmas with full metal and dynamically coated walls on EAST

We report the basic behaviors of fuel recycling and impurity accumulation during >100 s long-pulse H-mode plasma under full metal wall conditions on EAST. Significant fuel recycling and impurity rising, particularly from heavy impurities, have been observed when operating with a bare metal wall or a deteriorate real-time coated wall, which severely limits the duration of H-mode discharges. To address this issue, a novel dynamic wall coating technique combining feedforward and feedback controls has been successfully developed. The feedforward control presets the Li powder injection rate based on prior experimental observations, whereas the feedback control dynamically modulates the Li injection rate in response to real-time Li-II line emission measurements. Using this approach, a 605 s H-mode plasma has been achieved with fuel recycling and impurity level maintained stable. This result extends the previous record of a 403 s long-pulse H-mode plasma Gong et al (2024 Nucl. Fusion 64 112013) by over 200 s. It demonstrates the effectiveness of the dynamic powder injection technique in controlling fuel recycling and impurities accumulation, while prolonging plasma duration. These findings offer valuable insights into potential applications of other low-Z powder, such as boron, in ITER.

fuel recycling↗
DOE OSTI2021

Impurity screening behavior of the high-field side scrape-off layer in near-double-null configurations: prospect for mitigating plasma-material interactions on RF actuators and first-wall components

The impurity screening response of the high-field side (HFS) scrape-off layer (SOL) to localized nitrogen injection is investigated on Alcator C-Mod for magnetic equilibria spanning lower-single null, double-null and upper single null configurations under otherwise identical plasma conditions. L-mode, EDA H-mode and I-mode discharges are investigated. HFS impurity screening is found to depend on magnetic flux balance and the direction of B×?B relative to the most active divertor. Impurity ‘plume’ emission patterns indicate that both parallel and perpendicular (E×B) flows in the SOL contribute to the ‘flushing’ of impurities towards the active divertor, thereby affecting the overall impurity screening behavior. Despite the fact that the HFS scrape-off layer is extremely narrow in near-double-null configurations, this SOL is able to screen locally injected nitrogen at least as effectively as the low-field side (LFS) SOL – up to a factor of 10 more effective, depending on specific plasma conditions and whether the magnetic geometry produces parallel flows that work with or against E×B flows. For situations in which the E×B drift of the impurity ions opposes parallel flow toward the primary divertor, HFS impurity screening is found to be least effective. When E×B drifts assist parallel flow toward the primary divertor, HFS impurity screening is found to be very effective. These data support the idea of placing RF actuators and close-fitting wall components on the high-field side of the tokamak. With this configuration, near-double-null magnetic topologies may be used for active control of plasma parameters at the antenna/plasma interface for optimal RF coupling, mitigate the generation of local impurities from plasma-material interactions and, taking advantage of favorable plasma flows and good screening properties of the HFS SOL, further minimize the impact of wall-born impurity sources on the plasma core.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

MHD simulations of cold bubble formation from 2/1 tearing mode during massive gas injection in a tokamak

Massive gas injection (MGI) experiments have been carried out in many tokamaks to study disruption dynamics and mitigation schemes. Two events often observed in those experiments are the excitation of the m = 2, n = 1 magnetohydrodynamic mode, and the formation of cold bubble structure in the temperature distribution before the thermal quench (TQ). Here m is the poloidal mode number, n the toroidal mode number. The physics mechanisms underlying those phenomena, however, have not been entirely clear. In this work, our recent NIMROD simulations of the MGI process in a tokamak have reproduced the main features of both events, which has allowed us to examine and establish the causal relation between them. In these simulations, the 3/1 and 2/1 islands are found to form successively after the arrival of impurity ion cold front at the corresponding q = 3 and q = 2 rational surfaces. At the interface between impurity and plasma, a local thin current sheet forms due to an enhanced local pressure gradient and moves inward following the gas cold front, this may contribute to the formation of a dominant 2/1 mode. Following the growth of the 2/1 tearing mode, the impurity penetration into the core region inside the q = 2 surface gives rise to the formation of the cold bubble temperature structure and initiates the final TQ. Here, a subdominant 1/1 mode developed earlier near the q = 1 surface alone does not cause such a cold bubble formation, however, the exact manner of the preceding impurity penetration depends on the nature of the 1/1 mode: kink-tearing or quasi-interchange.

2/1 tearing mode↗
DOE OSTI2020

Electron Bernstein wave conversion of high-field side injected X-modes in QUEST

This paper presents a detailed design of the Q-shu University experimental steady state spherical tokamak's (QUEST's) high-field side (HFS) injection system for electron Bernstein wave (EBW) excitation and the results of an experimental comparison of the HFS eXtraordinary X-mode and low-field side (LFS) ordinary O-mode injection of 8.2 GHz radio frequency (RF) power. Waveguides, as an alternative to mirror polarizers for transmitting RF X-mode power from LFS to HFS for EBW conversion, were used instead of the installation of an RF mirror. Testing of LFS-to-HFS RF power transmission at 8.2 GHz, using an RG-50-type vacuum waveguide in a bench-scale device filled with SF6 gas at 0.03 Mpa, revealed that an RF power of 10.8 kW could traverse the fundamental electron cyclotron resonance layer for 60 s without breakdown. The short-length, open-ended waveguide antenna used in the HFS injection-induced wave diffraction reduced the efficiency of power delivery to the upper hybrid resonance (UHR) by approximately 7% at an electron temperature of 50 eV. The HFS injection was able to produce brighter camera images than the standard LFS injection. The location of the UHR, as estimated by measuring the density with an interferometer, agreed with its location as measured by plasma radiation low-field, side-edge positions shown by fast camera imaging. This indicates that the plasma was produced by mode-converted EBW. The HFS injection had an absorption efficiency of 96%, compared to 40% for LFS. A greater fluctuation of floating potential adjustable to the lower hybrid wave (LHW) was observed in the HFS case by installing a Langmuir probe, confirming that EBW conversion efficiency was higher in the HFS case. Moreover, after setting the poloidal field to B PF = 7.6 mT, plasma current (I P ) in the HFS peaked at 1.3 kA, as opposed to 0.3 kA for LFS, despite LFS injection having a total power of 55 kW, compared to 40 kW for HFS. However, as the impurity level was comparatively high, it is believed that this I P is dominated by pressure-drive, which makes it difficult to analyze EBWCD. Finally, the line-integrated density in the HFS injection peaked at 1.6 × 10 18 m -2 , compared to 8 × 10 17 m -2 in the LFS one.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Decoupling of Nitrogen and Oxygen Impurities in Nitrogen Doped SRF Cavities

The performance of superconducting radiofrequency (SRF) cavities is critical to enabling the next generation of efficient high-energy particle accelerators. Recent developments have focused on altering the surface impurity profile through \textit{in-situ} baking, furnace baking, and doping to introduce and diffuse beneficial impurities such as nitrogen, oxygen, and carbon. However, the precise role and properties of each impurity are not well understood. In this work, we attempt to disentangle the role of nitrogen and oxygen impurities through time-of-flight secondary ion mass spectrometry of niobium samples baked at temperatures varying from 75-800 $^\circ$C with and without nitrogen injection. From these results, we developed treatments recipe that decouple the effects of oxygen and nitrogen in doping treatments. Understanding how these impurities and their underlying mechanisms drive further optimization in the tailoring of impurity profiles for high-performance SRF cavities.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

The JOREK non-linear extended MHD code and applications to large-scale instabilities and their control in magnetically confined fusion plasmas

JOREK is a massively parallel fully implicit non-linear extended magneto-hydrodynamic (MHD) code for realistic tokamak X-point plasmas. It has become a widely used versatile simulation code for studying large-scale plasma instabilities and their control and is continuously developed in an international community with strong involvements in the European fusion research programme and ITER organization. This article gives a comprehensive overview of the physics models implemented, numerical methods applied for solving the equations and physics studies performed with the code. A dedicated section highlights some of the verification work done for the code. A hierarchy of different physics models is available including a free boundary and resistive wall extension and hybrid kinetic-fluid models. The code allows for flux-surface aligned iso-parametric finite element grids in single and double X-point plasmas which can be extended to the true physical walls and uses a robust fully implicit time stepping. Particular focus is laid on plasma edge and scrape-off layer (SOL) physics as well as disruption related phenomena. Among the key results obtained with JOREK regarding plasma edge and SOL, are deep insights into the dynamics of edge localized modes (ELMs), ELM cycles, and ELM control by resonant magnetic perturbations, pellet injection, as well as by vertical magnetic kicks. Also ELM free regimes, detachment physics, the generation and transport of impurities during an ELM, and electrostatic turbulence in the pedestal region are investigated. Regarding disruptions, the focus is on the dynamics of the thermal quench (TQ) and current quench triggered by massive gas injection and shattered pellet injection, runaway electron (RE) dynamics as well as the RE interaction with MHD modes, and vertical displacement events. Also the seeding and suppression of tearing modes (TMs), the dynamics of naturally occurring TQs triggered by locked modes, and radiative collapses are being studied.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Deuterium retention characteristics during lithium powder and granule injection in EAST

Fuel retention remains a critical challenge for magnetic confinement fusion devices. As a low-Z material, lithium plays a key role in tokamak wall conditioning and impurity control. In this study, the fuel retention behaviors associated with two lithium forms—powder and granule—are systematically examined using the gas balance method. Both forms of lithium injection significantly enhance fuel retention, shifting the wall behavior from net outgassing to net absorption. For lithium powder injection, compared with the reference discharge, a suppression efficiency exceeding 40% is achieved relative to the number of injected lithium atoms. Moreover, discharges with different injection rates show a monotonic increase in retained fuel with the injected lithium quantity. In contrast, repeated granule injections exhibit a pronounced cumulative effect, with the retention ratio varying from 0.16 to 0.65 over successive discharges, indicating progressive enhancement of wall absorption and deuterium retention. These results advance the understanding of wall behavior modification induced by solid material injection and provide insights for developing effective wall-conditioning strategies using low-Z materials in ITER and future fusion reactors, while the expected high retention may ultimately exclude the use of Li wall conditioning in future fusion devices.

Fuel retention↗
DOE OSTI2020

Observations of wall conditioning by means of boron powder injection in DIII-D H-mode plasmas

In this paper, we report observations from the DIII-D tokamak indicating that boron (B) powder injection in tokamak plasmas improves wall conditions similarly to glow discharge boronization (GDB). Isotopically enriched B powder (B 11 > 95%) was introduced gravitationally in a sequence of H-mode plasma discharges at rates up to ~160 mg s -1 for durations up to 3 s. Boron injection to cumulative amounts ≤0.1 g appeared to improve wall conditions similarly to boronization, with indications of reduced wall fueling, reduced recycling at the outer strike point and reduced impurity content at breakdown. Post-mortem analysis of graphite samples exposed to far scrape-off layer plasma fluxes during boron injection confirm the formation of a B-C layer, with average surface composition B:C ~ 1. The results suggest that injecting boron-rich powders in tokamak plasmas can effectively replenish boron films on carbon plasma facing components to improve wall conditions and extend the duration of the beneficial effects of GDB.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Global gyrokinetic simulations of electrostatic microturbulent transport in LHD stellarator with boron impurity

Abstract Global gyrokinetic simulations of electrostatic microturbulent transport for discharge # 166256 of the Large Helical Device stellarator in the presence of boron impurity show the co-existence of the ion temperature gradient (ITG) turbulence and trapped electron mode (TEM) turbulence before and during boron powder injection. ITG turbulence dominates in the core, whereas TEM dominates near the edge, consistent with the experimental observations. Linear TEM frequency increases from ∼ 80 kHz to ∼ 100 kHz during boron injection, and ITG frequency decreases from ∼ 20 kHz to ∼ 13 kHz, consistent with the experiments. The poloidal wave number spectrum is broad for both ITG (0–0.5 mm −1 ) and TEM (0–0.25 mm −1 ). The nonlinear simulations with boron impurity show a reduction in the heat conductivity compared to the case without boron. The comparison of the nonlinear transport before and during boron injection shows that the ion heat transport is substantially reduced in the region where the TEM is dominant. However, the average electron heat transport throughout the radial domain and the average ion heat transport in the region where the ITG is dominant are similar. The simulations with boron show the effective heat conductivity values qualitatively agree with the estimate obtained from the experiment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Direct internal recycling fractions approaching unity

Direct internal recycling (DIR) refers to the process of recovering pure hydrogen isotopes (D/T) from helium and other impurities in the fusion plasma exhaust and directing them back to the fuel injection system. Increasing the exhaust fraction purified through DIR significantly reduces the size and cost of the tritium plant and provides additional benefits including reduced requirements for both the tritium startup inventory and tritium breeding ratio. Metal foil pumps (MFPs) are the dominant technology for this separation, relying on the concept of superpermeation. We recently demonstrated that PdCu foils operated at low temperature provide both exceptional flux and resilience to helium absorption as the DIR fraction is increased. Herein we design and demonstrate continuous and semi-batch DIR processes using PdCu MFPs. Under continuous processing, stable performance was observed for DIR fractions up to 92 %. In addition, we demonstrate a semi-batch process capable of extending the DIR fraction to unity (> 99.8 %). Under the experimental conditions described within a PdCu MFP area of ~22 m 2 would be sufficient to process the fusion exhaust with 92 % DIR fraction at expected flowrates of 100 Pa·m 3 ·s -1 for a future fusion power plant.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Tungsten erosion and divertor leakage from the DIII-D SAS-VW tungsten-coated divertor in experiments with neon gas seeding

Collector probes have been used to examine tungsten divertor leakage in a variety of scenarios with low-Z impurity seeding during operation with the new tungsten-coated SAS-VW divertor in DIII-D. Measurements of tungsten deposition on collector probes inserted into the far Scrape-off-Layer (SOL) are used to deduce how efficiently tungsten leaks out of the closed, V-shaped divertor after it is eroded from the target surfaces. Qualitative differences in the tungsten deposition patterns across the collector probes provide clear experimental evidence that the SOL conditions depend on the low-Z impurity seeding conditions. These measurements show that in scenarios where neon gas is injected into the plasma, the tungsten divertor leakage and SOL transport depend on the poloidal location from which the neon is injected. In particular, neon injection from the Inner Midplane and Outer Midplane appear to each result in higher divertor leakage by a factor of 2 to 3 compared to cases with neon injection from either the SOL Crown or from the SAS-VW divertor itself.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

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

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

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Disruption thermal load mitigation with shattered pellet injection on the Joint European Torus (JET)

Disruption mitigation remains a critical, unresolved challenge for ITER. To aid in addressing this challenge, a shattered pellet injection (SPI) system was installed on JET and experiments conducted at a range of thermal energy fractions and stored energies in excess of 7 MJ. The primary goals of these experiments were to investigate the efficacy of the SPI on JET and the ability of the plasma to assimilate multiple pellets. Single pellet injections produced a saturation in total radiated energy (Wrad) with increasing injected neon content, suggesting total radiation of stored thermal energy. Further increases in injected neon quantities resulted in reduced cooling times and current quench (CQ) durations, indicating higher impurity assimilation. No significant variation in CQ duration or W rad was observed when varying the deuterium content at fixed neon quantities. Additionally, higher assimilation, inferred by shorter CQ durations, was measured when a mechanical punch was used to launch the pellets and this was attributed to a lower pellet velocity leading to higher solid content in the pellet plume and larger fragments penetrating deeper into the plasma. Radiation asymmetries averaged over the cooling time were inferred from Emis3D and ranged from 1.6 to 1.9. Asymmetries averaged over the entire disruption sequence were found to increase at higher thermal energy fractions. The radiated energy fractions decreased with increasing thermal energy fractions but this trend was eliminated when toroidal asymmetries were accounted for with Emis3D. Pure deuterium pellets were able to produce cooling times of up to 75 ms with a gradual loss in thermal stored energy of up to 80%. Experiments with multiple pellet injection indicated W rad can be increased through pellet superposition and density can be increased with an additional D2 injection without a reduction in W rad . KPRAD modelling accurately reproduced the cooling times and the CQ duration at high thermal energies. Assimilation estimates from KPRAD indicated CQ rates scale strongly whilst W rad scales weakly and saturates with assimilated neon content. Comparable W rad can be achieved with lower assimilated neon quantities as longer cooling times are attained. Thus reduced neon content can be preferential in a thermal load mitigation scheme as it may reduce radiation asymmetries and prevent flash melting.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Active wall conditioning through boron powder injection compatible $\mathrm{ELM}$ control in $\mathrm{EAST}$

Real-time wall conditioning and edge-localized mode (ELM) suppression through real-time boron powder injection into high-confinement mode discharges of the Experimental Advanced Superconducting Tokamak (EAST) have been investigated. The boron powder injection effectively conditioned the plasma-facing components, as shown by the reductions in low-Z and high-Z impurities and the global recycling level in the whole inner vessel, and suppressed the strong magnetohydrodynamic (MHD) activity in the plasma core. The improved wall conditions also resulted in an earlier transition from L- to H-mode. Boron powder injection also effectively suppressed ELM in the lower single null configuration with BT in the favorable direction while leaving the plasma density and stored energy largely unaffected.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Benign termination of runaway electron beams on ASDEX Upgrade and TCV

Abstract This paper discusses the development of a benign termination scenario for runaway electron (RE) beams on ASDEX Upgrade and TCV. A systematic study revealed that a low electron density (n e ) companion plasma was required to achieve a large MHD instability, which expelled the confined REs over a large wetted area and allowed for the conversion of magnetic energy to radiation. Control of the companion plasma n e was achieved via neutral pressure regulation and was agnostic to material injection method. The neutral pressure required for recombination was found to be dependent on impurity species, quantity and RE current. On TCV,n e increased at neutral pressures above 1 Pa, indicating that higher collisionality between the REs and neutrals may lead to an upper pressure limit. The conversion of magnetic energy to radiated energy was measured on both machines and a decrease in efficiency was observed at high neutral pressure on TCV. The benign termination technique was able to prevent any significant increase in maximum heat flux on AUG from 200 to 600 kA of RE current, highlighting the ability of this approach to handle fully formed RE beams.

Physics↗
DOE OSTI2025

Wall conditioning effects of boron powder injection in KSTAR with a tungsten divertor

Boron powder dropped into KSTAR plasmas decreased the radiated power, core electron density, and Z eff , indicating that the ablated and redeposited boron powder helped to condition the plasma-facing surfaces. Visible line emission of oxygen and tungsten were also reduced by 50% (80%) with boron injection into H-mode (L-mode) discharges that used the new KSTAR tungsten monoblock lower divertor. Dynamic particle balance analysis found a negligible difference in the inferred wall pumping rate during the steady portion of the discharges. It is inferred that the observed conditioning effects were principally caused by a reduction in intrinsic impurities as opposed to a reduction in wall recycling. These results are in qualitative agreement with low-Z injected powder experiments across many fusion devices, confirming the utility of low-Z powder injection as a real-time wall conditioning tool.

boron powder injection↗