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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Role of E × B Drift in Divertor Detachment Control via Boron Powder Injection on EAST

Here, the effects of B powder injection on plasma detachment about EAST discharge were studied by using SOLPS-ITER code package with the effects of E × B drifts considered. The simulation results show that plasma detachment occurs at the inner target in favourable toroidal magnetic field (B t ) direction at a relatively low B powder flow rate, one order of magnitude lower than that at the outer target. In a similar scenario with unfavourable B t , it is found that the detachment thresholds of B flow rate for both the inner and outer targets are close and of the same order as that for the outer target with favourable Bt. In favourable B t direction at B powder flow rate of 1.2 × 10 21 atoms/s, a localized, broadened high-density region is formed near the inner target benefitted by the injection location and the E × B drift, and a radiation-intensified zone, mostly contributed by B 1+ and B 2+ , occurs there. The E × B drift facilitates plasma detachment at the inner target and simultaneously amplifies the in–out divertor asymmetry. In addition, the simulation results with three different injection locations show that the injection from outer strike point leads to the lowest Z eff inside the separatrix and has an intermediate flow rate for detachment at the outer target, comparing with the X-point and upstream locations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Integration of high confinement, high poloidal beta plasma with dual radiated power and detachment controls for divertor protection and ELM suppression

Divertor detachment without serious core confinement quality loss in DIII-D’s high poloidal β scenario has been combined with impurity-induced ELM mitigation without disruption. Use of Ne previously granted access to a detached, non-ELMing regime that retained high confinement quality due to stimulation of Internal Transport Barrier (ITB) growth, but suffered from on-going core fuel dilution and high disruptivity. Excess Ne accumulation in the core plasma has now been avoided by feeding back core radiated power ($P_{rad}$) measurements to control Ne seeding, rather than using attachment fraction ($A_{frac}$) control with Ne; this also reduces disruptivity. At the same time, N 2 seeding is used in a feedback loop with $A_{frac}$ measurements, which previously posed low disruption risk. In this way, the effect of Ne in the core is managed while avoiding excess seeding, and N 2 acts to correct for any excess heat exhaust that might interfere with detachment. The average Ne flow rate was 38% of what was used in pure Ne $A_{frac}$ control, plus average N 2 flow that was 43% of pure N 2 $A_{frac}$ control, even while meeting an even deeper detachment target. Meanwhile, a steep pressure gradient in the core plasma reduces sensitivity to impurity-induced degradation of the pedestal and reasonable confinement quality was maintained despite operational challenges that blocked formation of an ITB in these experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Recrystallization, cracking, and erosion of dispersoid-strengthened tungsten materials during exposure to divertor plasmas

In this study, we investigated the effects of combined intense particle and heat flux exposure on advanced tungsten plasma-facing materials within the DIII-D fusion facility. Our test matrix included two types of dispersoid-strengthened tungsten (containing either 100 nm diameter TiO 2 or Ni particles), along with high-purity polycrystalline tungsten as a reference. This experiment relied on a sample geometry angled at 15° relative to the divertor surface, thereby allowing the surfaces to intercept steady-state perpendicular heat fluxes (q ⟂ ) ranging from 10.1 to 19.6 MW/m 2 . During each shot, the samples were exposed to 42 Hz edge-localized modes (ELMs), allowing us to test the material response to transient heating. We correlated the exposure conditions with extensive post-test surface composition analysis and microscopy to determine how the plasma modified each surface. The angled specimens closest to the strike point received the highest combined heat and particle flux and melted midway through the experiment. EBSD analysis revealed they were completely recrystallized throughout, with an average grain size >100 µm. On the other hand, the specimens that received a lower steady state heat flux survived with more superficial surface damage. Whereas the high-purity polycrystalline tungsten exhibited a higher surface roughness, the dispersoid-strengthened material exhibited more extensive shallow inter-granular cracking. In addition, the surface was depleted of dispersoids following plasma exposure, possibly because of evaporation and/or sputtering. The results described here provide insights into the performance of these materials in a fusion environment which can guide further optimization for use in long-pulse devices.

Kolasinski, Robert D. [Sandia National Laboratorie↗
DOE OSTI2021

Simple, general, realistic, robust, analytic tokamak equilibria. Part 1. Limiter and divertor tokamaks

Tokamak equilibria have been derived that are analytic solutions to the Grad–Shafranov equation. Here in this paper, Part 1, describes a wide range of such equilibria including smooth limiter surfaces, double- and single-null divertor surfaces, arbitrary aspect ratio, elongation, triangularity and beta. Part 2 generalizes the analysis to include edge pedestals and toroidal flow.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Enhanced particle flux due to localized divertor MHD instability in DIII-D tokamak

An MHD mode with a frequency of <10 kHz has been identified near the inner strike point from various diagnostics, i.e., divertor Langmuir probes, magnetics sensors, and interferometers, but does not appear in the upstream and core diagnostics. This MHD mode is associated with magnetic oscillations of ≳5 G, has a long wavelength in the toroidal direction with toroidal mode number n=1, but is localized in a narrow radial region of a few cm. The mode appears when the outer strike point remains attached and the inner strike point nearly detaches, grows with increasing density, and eventually weakens and vanishes as the outer strike point detaches. This mode results in particle flux with an order of magnitude higher than the background plasmas near the inner strike point. The mode characteristics are consistent with the Current-Convective-Instability theory prediction. Initial simulations based on experimental input have found oscillations with similar frequencies but weaker amplitude.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2020

The Langmuir probe system in the Wendelstein 7-X test divertor

Here, the design and evaluation of the Langmuir probe system used in the first divertor operation phase of Wendelstein 7-X is described. The probes are integrated into the target plates and have individually facetted surfaces to keep the angle of incidence of the magnetic field within an appropriate range for different magnetic configurations. Multiple models for the derivation of plasma parameters from current–voltage characteristics are introduced. These are analyzed with regard to their assumptions and limitations, generalized, and adapted to our use case. A detailed comparison is made to determine the most suitable model. It is found that the choice of model has a large impact, for example, resulting in a change in the inferred temperatures of up to a factor two. This evaluation is implemented in a Bayesian modeling framework and automated to allow for joint analysis with other diagnostics and a replacement of ad hoc assumptions. We rigorously treat parameter uncertainties, revealing strong correlations between them. General and flexible model formulations permit an expansion to additional effects.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Density drop at the divertor target in the prototype material plasma exposure eXperiment (Proto-MPEX)

The steady-state linear device “Material Plasma Exposure eXperiment” (MPEX) is currently under construction at Oak Ridge National Laboratory with the goal of enabling Plasma-Material Interaction studies at future fusion reactor relevant plasma conditions. In this work, a newly in-house developed hybrid Particle-In-Cell code-PICOS++ is applied to understand the experimental results obtained from the prototype of MPEX referred to as the “Proto-MPEX” during its helicon-only and helicon with ion cyclotron resonance heating (ICRH) experiments. This study explains the physics of the experimentally observed plasma density-drop at the divertor target in Proto-MPEX device during ICRH. In contrast to previous work on ICRH in MPEX, this study demonstrates that the mirror force plays a central role in the Proto-MPEX plasma transport during ICRH, which has new features not previously explored. Force balance analyses reveal that the temperature anisotropy produced by ICRH leads to a significant increase in the mirror force downstream of the resonance where the magnetic field is diverging. This force accelerates ions toward the target and leads to a drop in plasma density to ensure conservation of particle flux. Simulations with ICRH where the magnetic field divergence downstream of the resonance has been removed, do not produce plasma acceleration nor density drop at the target despite efficient ion heating at the resonance. Moreover, simulation results demonstrate that for a given ICRH power, lowering the source rate produces ions with increased perpendicular energy which interact with the mirror force to produce higher plasma acceleration which increases the strength of the density-drop at the target. The strength of the density drop appears to reach an asymptotic limit at a certain threshold ICRH power. Simulations show that this threshold power increases with increasing particle source rate.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

ORBIT simulations of fast ion power loads on the wall of the Divertor Tokamak Test

Abstract Neutral beam injection in tokamaks produces a population of fast ions, which interact with 3D magnetic fields in a variety of ways, often resulting in energetic particle losses in very short times to the wall. Careful design of neutral beams and active control of error fields helps to keep these losses to a minimum. Nevertheless, past experience in tokamaks in the 1980s and detailed simulations for future machines, such as International Thermonuclear Experimental Reactor, suggest that very localized fast ion losses (‘hot spots’) can be present, even if the overall losses are low. In this paper, we discuss this issue in the Divertor Tokamak Test (DTT) project, and we show that in the standard single-null full-power scenario of the DTT, fast ions produce two hot spots, corresponding approximately to the beam injection and exit toroidal angles: the former being mainly due to prompt losses/passing particles, while the latter is due to trapped ions. However, the maximum power load in these spots is of the order ∼ 63 kW m −2 , below the tolerance of plasma-facing components of the machine.

Physics↗
DOE OSTI2025

Electron transport in a tokamak scrape-off layer: impact of toroidal nonuniformities of divertor targets

An analysis of electron dynamics is carried out in the tokamak scrape-off layer (SOL) region. Small non-axisymmetric perturbations of the divertor target sheath potential affect electron drift orbits in the SOL and may lead to non-ambipolar electron radial transport. The resulting rates of electron convection, diffusion, and mobility are calculated analytically, and the analytic results are compared with direct numerical simulations of electron drift orbits. The proposed mechanism of non-ambipolar electron transport may be relevant to the sustainment of the SOL plasma quasi-neutrality in the ‘heuristic’ model of the SOL width (Eich et al 2011 Phys. Rev. Lett. 107 215001).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Comparative study of boron and neon injections on divertor heat fluxes using SOLPS-ITER simulations

Here, based on the EAST equilibrium, the effects of boron (B) and neon (Ne) injected at different locations on the target heat load, and the distributions of B and Ne particles were investigated by transport code SOLPS-ITER. It was found that the B injection was more sensitive to the injection location for heat flux control than impurity Ne. The high electron and ion densities near the inner target in the discharge with impurity B injected from over X-point (R 1 ) led to plasma detachment only at the inner target, and the localized B ions in the cases with injection from outer target location (R 2 ) and upstream location (R 3 ) led to far-SOL detachment at the outer target, but not at the inner target. In contrast, for Ne, the spatial distributions of Ne ions and electrons were found to be similar in all the cases at the three injection locations, and the detached plasma was achieved at the inner target and the electron temperature was reduced at the outer target. For locations R 2 and R 3 , impurity B showed a more pronounced effect on the heat flux at the far-SOL of the outer target. Further analysis indicated that Ne atoms came mainly from the recycling sources, whereas B atoms came mainly from injection, and that their distinct atomic distributions resulted from the difference in the ionization threshold and ionization mean free path. In addition, the radiation proportion of B in the divertor region was larger than that of Ne when the total radiation power was similar, which suggests that B has less influence on the core region.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

New heat flux model for non-axisymmetric divertor infrared structures

A convective heat flux model for perturbed plasmas, based on guiding center ion drift in vacuum elds (A. Wingen, et al., Phys. Plasmas 21 (2014) 012509), has been updated. The old model only considered ion heat flux, while here also electron heat flux is included. The updated model predicts divertor heat flux distributions in non-axisymmetric (3D) plasmas with applied Resonant Magnetic Perturbation (RMP) fields, and includes electric scalar potentials. It is found that a radial electric field in the near Scrape-o Layer (SOL) can considerably shift the footprints toroidally, leading to a smearing out eff ect of the incident heat flux, while a simple model for sheath potential has little impact on footprints. Various approaches to model electron heat flux are studied. A convective electron model, based on collisionless free streaming, is found to yield the best agreement with measurements, while a conductive model requires a at temperature gradient inside lobes to yield acceptable peak heat ux values. A heuristic heat flux layer approach, based on a fixed layer width also requires a limited heat flux inside the last closed flux surface (LCFS); by selecting various locations of the LCFS, the results of the conductive or convective model can be recovered respectively. The sum of ion and electron heat fluxes, both obtained by the convective model, is compared to experimental data for multiple time slices in DIII-D. Strike point splitting is observed with peak heat fluxes and layer widths that compare well to infrared camera (IR) measurements.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Collisionless losses of fast ions in the Divertor Tokamak Test due to toroidal field ripple

In this paper we analyze fast ion motion in the Divertor Tokamak Test device. DTT is planned to be heated through a mix of 45 MW heating power, including 15 MW negative-ion-based Neutral Beam heating (NNBI) which is currently being developed by Consorzio RFX in Padova, Italy. An issue for DTT is that a Toroidal Field (TF) ripple with a maximum value of about 0.42% (with respect to the on-axis magnetic field $B_0$) is expected on the low-field side (LFS), and this ripple interacts with fast ions through rather well-known phenomena of ripple-precession resonances, in addition to prompt losses of ions which do not complete a full orbit in the poloidal plane. We will show that, with the planned geometry of NNBI, prompt losses are negligible, and ripple-precession losses amount to a maximum 0.15%. The calculations are performed with the guiding center code ORBIT using two different equilibria, and a beam with energy 400 keV and injection angle $\alpha_{inj} = 40^{\circ}$ (measured w.r.t. the first wall), which corresponds to a pitch of injected particles $\lambda = v_{\parallel}/v \approx \sin{\alpha_{inj}} = 0.65$. Main resonances are of the form $\omega_b - n N \omega_d = 0$, $\omega_b$ and $\omega_d$ being the bounce and precession frequency, respectively, $N=18$ the ripple periodicity and $3 \le n \le 6$ are the toroidal wavenumbers of the resonances. Although collisionless interaction with the TF ripple does not pose a serious threat to the NNBI project, an open question remains whether the presence of these resonances will interact with fast particles accelerated by \alfven eigenmodes, and if stochastization of the resonances is possible in DTT, as it was observed in the past in TORE SUPRA.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Prediction of divertor heat flux width for ITER pre-fusion power operation using BOUT++ transport code

Prediction of divertor heat flux width is performed for the first and the second pre-fusion power operation (PFPO) phases specified in the new ITER research plan using BOUT++ transport code (Li et al 2018 Comput. Phys. Commun.228 69–82). Here the initial plasma profiles inside the separatrix are taken from CORSICA scenario studies. Transport coefficients in transport code are calculated by inverting the plasma profiles inside the separatrix and are assumed to be constants in the scrape-off-layer. An anomalous thermal diffusivity scan is performed with E × B and magnetic drifts. The results in two scenarios identifying two distinct regimes: a drift-dominant regime when diffusivity is smaller than the respective critical thermal diffusivity χ c and a turbulence-dominant regime when diffusivity is larger than it. The Goldston heuristic drift model and the ITPA multi-machine experimental scaling yield a lower limit of the width λ q . From transport simulations, we obtain the critical thermal diffusivity χ c = 0.5 m 2 s –1 for the PFPO-1 scenario with toroidal magnetic field B = 1.77 T and plasma current I p = 5 MA, and χ c = 0.3 m 2 s –1 for the PFPO-2 scenario with toroidal magnetic field B = 2.65 T and plasma current I p = 7.5 MA. Separatrix temperature and collisionality also have a significant impact on the heat flux width in the drift-dominant regime. The investigation clearly yields a scaling for critical thermal diffusivity ${\chi }_{\text{c}}\propto {A}^{1/2}/(Z{\left(1+Z\right)}^{\frac{1}{2}}{B}_{\text{p}}^{2})$ using ITER scenarios with fixed safety factor q 95 , major radius R, aspect ratio R/a, and the separatrix temperature T sep , and establishes the connection with CFETR and C-Mod discharges. This scaling implies that for a given tokamak device with q 95 , R, R/a, and T sep fixed, a reduction of poloidal magnetic field by a factor of 3 leads to a 9 times higher critical value of thermal diffusivity χ c , possibly yielding a transition from turbulence- to drift-dominant regime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Divertorlets concept for low-recycling fusion reactor divertor: experimental, analytical and numerical verification

The 'divertorlets' concept is a potential non-evaporative liquid metal solution for heat removal at low recycling regime. A toroidal divertorlets prototype was built and tested in LMX-U at Princeton Plasma Physics Laboratory to evaluate the performance of this configuration. Here, details of the design, experimental results, comparison with analytical theory and MHD numerical simulations of toroidal divertorlets are covered. Experiments, analytical model and simulations showed agreement and allowed the projection of operation properties at higher magnetic flux densities (reactor-like operation), proving the concept to be a compelling solution for divertor applications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Characteristics of grassy ELMs and their impact on the divertor heat flux width

We report BOUT++ turbulence simulations are conducted for a 60 s steady-state long pulse high β p EAST grassy ELM discharge. BOUT++ linear simulations show that the unstable mode spectrum covers a range of toroidal mode numbers from low-n (n = 10–15) peeling–ballooning modes (P–B) to high-n (n = 40–80) drift-Alfvén instabilities. Nonlinear simulations show that the ELM crash is triggered by low-n peeling modes and fluctuation is generated at the peak pressure gradient position and radially spread outward into the scrape-off-layer, even though the drift-Alfvén instabilities dominate the linear growth phase. However, drift-Alfvén turbulence delays the onset of the grassy ELM and enhances the energy loss with the fluctuation extending to pedestal top region. Simulations further show that if the peeling drive is removed, the fluctuation amplitude drops by an order of magnitude and the ELM crashes disappear. The divertor heat flux width is ~2 times larger than the estimates based on the HD model and the Eich's ITPA multi-tokamak scaling (or empirical Eich scaling) due to the strong radial turbulence transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

L-H transition studies in tritium and deuterium–tritium campaigns at JET with Be wall and W divertor

The recent deuterium–tritium campaign in JET-ILW (DTE2) has provided a unique opportunity to study the isotope dependence of the L-H power threshold in an ITER-like wall environment (Be wall and W divertor). Here we present results from dedicated L-H transition experiments at JET-ILW, documenting the power threshold in tritium and deuterium–tritium plasmas, comparing them with the matching deuterium and hydrogen datasets. From earlier experiments in JET-ILW it is known that as plasma isotopic composition changes from deuterium, through varying deuterium/hydrogen concentrations, to pure hydrogen, the value of the line averaged density at which the threshold is minimum, ${\bar n_{{\text{e}},{\text{min}}}}$, increases, leading us to expect that ${\bar n_{{\text{e}},{\text{min}}}}$(T) < ${\bar n_{{\text{e}},{\text{min}}}}$(DT) < ${\bar n_{{\text{e}},{\text{min}}}}$(D) < ${\bar n_{{\text{e}},{\text{min}}}}$(H). The new power threshold data confirms these expectations in most cases, with the corresponding ordering of the minimum power thresholds. We present a comparison of this data to power threshold scalings, used for extrapolation to future devices such as ITER and DEMO.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Neural networks for estimation of divertor conditions in DIII-D using C III imaging

Deep learning approaches have been applied to images of C III emission in the lower divertor of DIII-D to develop models for estimating the level of detachment and magnetic configuration (X-point location and strike point radial location). The poloidal distance from the target to the C III emission front is used to represent the level of detachment. The models perform well on a test dataset not used in training, achieving $F_1$ scores as high as 0.99 for detachment state classification and root mean squared error (RMSE) as low as 2cm for front location regression. Predictions for shots with intermittent reattachment are studied, with class activation mapping used to aid in interpretation of the model predictions. Based on the success of these models, a third model was trained to predict the X-point location and strike point radial position from C III images. Though the dataset covers only a small range of possible magnetic configurations, the model shows promising results, achieving RMSE around 1cm for the test data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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↗