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At least 451 records · Page 25

Density limits as disruption forecasters for spherical tokamaks

Fusion power output from spherical tokamaks would benefit from increased confined plasma density, but there exists a limit on the density before confinement is lost and the plasma current is disrupted. This density limit has long been characterized by a simple, global Greenwald limit proportional to the plasma current and inversely proportional to the cross sectional area of the plasma. It is shown that in the database of discharges from the National Spherical Tokamak Experiment (NSTX) and Mega Ampere Spherical Tokamak (MAST) spherical tokamaks, the likelihood of disruption does increase above the Greenwald limit, and especially in the plasma current rampdown phase. The physics of the density limit has been recently theoretically explored through local criteria. Several of these are tested using the disruption event characterization and forecasting (DECAF TM ) code for their potential effectiveness as disruption warning signals. For a limited set of NSTX discharges, a local island power balance criteria was found to be less reliable, presently, than the Greenwald limit. For this work, an empirical critical edge line density and a boundary turbulent transport limit were both tested for MAST-U, which has an electron density profile measurement with high spatial resolution in the outer part of the plasma. Both were found to have similar dependencies on key plasma parameters. In a limited set of MAST-U discharges that appear to disrupt due to rising density at values under the Greenwald limit, crossing of the boundary turbulent transport limit occurred close to the time of disruption. Finally, these limits were evaluated for their potential use in real-time, and it was found that with the necessary real-time inputs and with refinement through further testing, these limits could be implemented in a real-time disruption forecasting system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Design and engineering challenges of a multi-energy hard x-ray camera for long-pulse profile measurements at WEST tokamak

The WEST tokamak has recently been prepared for long pulse operation with a water-cooled full first wall. Heating is provided by radiofrequency systems, including Lower Hybrid Current Drive (LHCD). The Princeton Plasma Physics Laboratory has developed a multi-energy hard x-ray camera for profile measurements at WEST. The camera is based on a new generation of 2D pixel array detectors that allow the energy threshold to be independently set for each pixel. The diagnostic will provide spatial, temporal and energy resolved measurements of the hard x-ray emission from the full plasma cross-section, investigating several physical quantities such as the electron temperature from continuum emission, the fast electron tail density produced by radiofrequency current drive and runaway electrons, as well as characteristic tungsten x-rays due to beam-target emission at the edge. Overall, this work describes the engineering challenges that the WEST long pulse scenario poses for this diagnostic and how the design addresses and solves them. Vacuum, thermal stress and heat transfer calculations are presented and discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Influence of triangularity on the plasma response to resonant magnetic perturbations

The effect of the axisymmetric plasma shape on the non-axisymmetric plasma response to resonant magnetic perturbations is investigated in experiment and modeling for the DIII-D, EAST and ASDEX Upgrade tokamaks. Systematically modeling the effect of the triangularity, whilst keeping other equilibrium quantities largely unchanged, reveals that the plasma response is strongly suppressed at high triangularity compared to that at low triangularity. This is validated through targeted comparison with experiments at DIII-D, EAST, and ASDEX Upgrade. DIII-D and EAST magnetic measurements are used to validate simulations, while at ASDEX Upgrade the plasma edge displacement is measured. Both experiments and modeling find a reduced magnetic plasma response on the high-field side at high triangularity across devices. Multi-modal analysis of the simulation results extracts the mode structure and applied spectrum dependence of each mode. The amplitude of the dominant mode reveals similar trends with the edge resonance and radial displacement near the X-point, which suggests that the multi-mode plasma response provides another way to understand the edge localized mode (ELM) control physics. The plasma response is strongly reduced at high triangularity compared to that at low triangularity, which implies different ELM control effects as shaping is varied. Furthermore, these findings indicate that the plasma shape should be taken into consideration when designing an RMP-ELM control strategy in experiment, and that predictive plasma response calculations can be used to maximize access to RMP-ELM control in future devices by maximizing the coupling between coils and the plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Design of a Thomson scattering diagnostic for the SMall Aspect Ratio Tokamak (SMART)

Here, we describe the design of a Thomson scattering (TS) diagnostic to be used on the SMall Aspect Ratio Tokamak (SMART). SMART is a spherical tokamak being commissioned in Spain that aims to explore positive triangularity and negative triangularity plasma scenarios at a low aspect ratio. The SMART TS diagnostic is designed to operate at high spatial resolution, 6 mm scattering length in the low-field side and 9 mm in the high-field side regions, and a wide dynamic range, electron temperature from 1 eV to 1 keV and density from 5 x 10 18 m -3 to 1 x 10 20 m -3 ⁠, to resolve large gradients formed at the plasma edge and in the scrape-off layer (SOL) under different triangularities and low aspect ratios. A 2 J @1064 nm laser will be used that is capable of operating in the burst mode at 1, 2, and 4 kHz to investigate fast phenomena and at 30 Hz to study 1 s (or more) long discharges. The scattered light will be collected over an angular range of 60° – 120° from 28 spatial points in the midplane covering the entire plasma width and the outer midplane SOL. Each scattering signal will be spectrally resolved on five wavelength channels of a polychromator to obtain the electron temperature measurement. We will also present a method to monitor in situ laser alignment in the core during calibrations and plasma operations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Transport-driven toroidal rotation with general viscosity profile

Abstract Using the assumption of a weak normalized turbulent viscosity, usually valid in practice, the modulated-transport model (Stoltzfus-Dueck 2012 Phys. Plasmas 19 055908) is generalized to allow the turbulent transport coefficient to vary in an arbitrary way on radial and poloidal position. The new approach clarifies the physical interpretation of the earlier results and significantly simplifies the calculation, via a boundary-layer asymptotic method. Rigorous detailed appendices verify the result of the simple boundary-layer calculation, also demonstrating that it achieves the claimed order of accuracy and providing a concrete prediction for the strong plasma flows in the immediate vicinity of the last closed flux surface. The new formulas are used to predict plasma rotation at the core-edge boundary, in cases with and without externally applied torque. Dimensional formulas and extensive discussion are provided, to support experimental application of the new model.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Fast-ion physics in SPARC

Potential loss of energetic ions including alphas and radio-frequency tail ions due to classical orbit effects and magnetohydrodynamic instabilities (MHD) are central physics issues in the design and experimental physics programme of the SPARC tokamak. The expected loss of fusion alpha power due to ripple-induced transport is computed for the SPARC tokamak design by the ASCOT and SPIRAL orbit-simulation codes, to assess the expected surface heating of plasma-facing components. We find good agreement between the ASCOT and SPIRAL simulation results not only in integrated quantities (fraction of alpha power loss) but also in the spatial, temporal and pitch-angle dependence of the losses. If the toroidal field (TF) coils are well-aligned, the SPARC edge ripple is small (0.15–0.30 %), the computed ripple-induced alpha power loss is small ( ~0.25% ) and the corresponding peak surface power density is acceptable ( 244 kW m -2 ). However, the ripple and ripple-induced losses increase strongly if the TF coils are assumed to suffer increasing magnitudes of misalignment. Surface heat loads may become problematic if the TF coil misalignment approaches the centimetre level. Ripple-induced losses of the energetic ion tail driven by ion cyclotron range of frequency (ICRF) heating are not expected to generate significant wall or limiter heating in the nominal SPARC plasma scenario. Because the expected classical fast-ion losses are small, SPARC will be able to observe and study fast-ion redistribution due to MHD including sawteeth and Alfvén eigenmodes (AEs). SPARC's parameter space for AE physics even at moderate Q is shown to reasonably overlap that of the demonstration power plant ARC (Sorbom et al., Fusion Engng Des., vol. 100, 2015, p. 378), and thus measurements of AE mode amplitude, spectrum and associated fast-ion transport in SPARC would provide relevant guidance about AE behaviour expected in ARC.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Understanding the negative triangularity ELM trigger and ELM free state on DIII-D with ECE-imaging

Here, the Electron Cyclotron Emission Imaging (ECEI) diagnostic was used to observe a finite-n interchange mode structure in the edge of negative triangularity shaped plasmas on DIII-D. At a small negative triangularity (δ u = -0.2), the plasma is in the H-mode with ELMs that are triggered by a low-n interchange mode. At a larger negative triangularity (δ u = -0.4) and low NBI power (2 MW), a dithering oscillation is observed that is triggered by a low-n interchange mode, whereas at higher NBI power (>2 MW), the edge reverts to L-mode and the low-n interchange mode is present continuously. In all cases, the edge pressure gradient is clamped when the interchange mode is present. It is concluded that the low-n interchange mode prevents the plasma from transitioning to H-mode at a large negative triangularity. This agrees with linear BOUT++ simulations which suggest that the interchange-type MHD can be a resistive ballooning mode whereby resistivity can significantly increase the finite-n ballooning mode growth rate. The absence of H-mode at large negative triangularity can, thus, be explained by the excitation of low-n pressure driven resistive ballooning modes in the plasma edge.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

The Diagnostics of the External Plasma for the Plasma Rocket

The plasma rocket is located at NASA Johnson Space Center. To produce a thrust in space. an inert gas is ionized into a plasma and heated in the linear section of a tokamak fusion device to 1 x 10(exp 4) - 1.16 x 10(exp 6)K(p= 10(exp 10) - 10(exp 14)/cu cm ). The magnetic field used to contain the plasma has a magnitude of 2 - 10k Gauss. The plasma plume has a variable thrust and specific impulse. A high temperature retarding potential analyzer (RPA) is being developed to characterize the plasma in the plume and at the edge of the magnetically contained plasma. The RPA measures the energy and density of ions or electrons entering into its solid angle of collection. An oscilloscope displays the ion flux versus the collected current. All measurements are made relative to the facility ground. A RPA is being developed in a process which involves the investigation of several prototypes. The first prototype has been tested on a thermal plasma. The knowledge gained from its development and testing were applied to the development of a RPA for collimated plasma. The prototypes consist of four equally spaced grids and an ion collector. The outermost grid is a ground. The second grid acts as a bias to repel electrons. The third is a variable v voltage ion suppressor. Grid four (inner grid) acts to repel secondary electrons, being biased equal to the first. Knowledge gained during these two stages are being applied to the development of a high temperature RPA Testing of this device involves the determination of its output parameters. sensitivity, and responses to a wide range of energies and densities. Each grid will be tested individually by changing only its voltage and observing the output from the RPA. To verify that the RPA is providing proper output. it is compared to the output from a Langmuir or Faraday probe.

Karr, Gerald R.↗

3D ion gyro-orbit heat load predictions for NSTX-U

Abstract High power tokamaks operate with divertor heat loads capable of destroying the plasma facing components (PFCs). High fidelity heat load predictions are necessary to ascertain the PFC state for design and during operation. Typical heat flux calculations are 2D, time invariant, and assume that power flows directly along the magnetic field lines (the optical approximation). These assumptions neglect the complex 3D geometries employed to protect the PFCs, the time varying nature of the plasma and PFC thermal state, and the helical trajectories of ions with finite Larmor radii (the gyro-orbit approximation). An integrated software framework, the heat flux engineering analysis toolkit (HEAT), was developed to generate time varying optical heat loads applied to real engineering computer aided design (CAD) (Looby et al 2022 Fusion Sci. Technol. 78 10–27). Recently, an ion-gyro orbit module has been added to HEAT. This module calculates the helical trajectories of ions as they gyrate about the magnetic field lines using kinetic theory macro-particles to accelerate the calculation. First, the new gyro-orbit module will be presented. Next, a comparison to existing research is performed. Finally, an analysis of the gyro-orbit heat loads for NSTX-U is presented for diverted discharges using the engineering CAD models utilized for PFC fabrication. Including these gyro-orbit effects can enhance the PFC performance by ‘smearing’ out the magnetic shadows associated with the castellated fish-scaled geometry. Simultaneously, the helical trajectories can degrade performance when they load narrow regions on edges and corners with high heat fluxes. Analysis of the trade-offs between these competing effects is included, and regions for further investigation are identified.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Effects of density turbulence on helicon wave propagation in the core plasmas

Radio frequency wave propagation can be significantly affected by density irregularities, such as filaments in the scrape-off layer or instabilities in the core plasma. In this study, we examine the impact of edge turbulence on helicon wave propagation using the Petra-M simulation tool. To analyze the effect of edge turbulence, we utilize a realistic background plasma derived from XGC simulations, which includes spatial density fluctuations at the edge. This focus is particularly relevant for the DIII-D configuration, characterized by edge density turbulence of the core plasmas in a wide pedestal QH-mode. We focus on helicon wave propagation in the core plasma since the slow mode cannot propagate into the core plasma when the density is higher than the lower hybrid resonance, where a wide pedestal QH-mode occurs. The simulation results indicate that edge density fluctuations have a substantial impact on helicon wave coupling due to scattering in the core. We specifically demonstrate that the toroidal mode number and the level of density fluctuations are significant factors influencing scattering. While we have minimized the excitation of slow modes from the antenna, we still observe mode-converted slow modes resulting from the incoming helicon waves. The insights derived from these simulations will inform upcoming tokamak experiments regarding helicon antenna coupling in long pulse scenarios.

Kim, Eun-Hwa [Princeton Plasma Physics Laboratory ↗

Improved core-edge compatibility using impurity seeding in the small angle slot (SAS) divertor at DIII-D

Impurity seeding studies in the slot divertor (SAS) at DIII-D have revealed a strong relationship between the detachment onset and pedestal characteristics with both target geometry and impurity species. N 2 seeding in the slot has led to the first simultaneous observation of detachment on the entire suite of boundary diagnostics viewing the SAS without degradation of core confinement. SOLPS-ITER simulations with D+C+N, full cross-field drifts and n-n collisions activated are performed for the first time in DIII-D to interpret the behavior. This illustrates a strong effect of divertor configuration and plasma drifts on the recycling source distribution with significant consequences on plasma flows. Flow reversal is found for both main ions and impurities affecting strongly the impurity transport and providing an explanation for the observed dependence on strike point location of the detachment onset and impurity leakage found in the experiments. Matched discharges with either nitrogen or neon injection show that while nitrogen does not significantly affect the pedestal, neon leads to increased pedestal pressure gradient and improved pedestal stability. Little nitrogen penetrates in the core, but a great amount of neon is found in the pedestal consistent with the different ionization potentials of the two impurities. This research demonstrates that neutral and impurity distributions in the divertor can be controlled through variations in strike point locations in a fixed baffle structure. Divertor geometry combined with impurity seeding enables mitigated divertor heat flux balancing core contamination and thus leading to enhanced divertor dissipation and improved core-edge compatibility which are essential for ITER and for future fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Turbulence characteristics and flow dynamics impacts on the H-mode transition in favourable magnetic geometry with lower power threshold

We report the L-H transition power threshold (P LH ) in favourable magnetic geometry (ion ∇B drift pointing towards X-point) is much lower than in the unfavourable magnetic geometry (ion ∇B drift pointing away from X-point) on multiple tokamaks. In a systematic experiment on DIII-D, the ion ∇B drift direction was changed continuously from the unfavourable to favourable configuration during plasma discharges. During such process, the input neutral beam power was kept constant at a value that was above P LH for favourable configuration, but lower than P LH for unfavourable configuration. Toroidal field and plasma current were also kept constant and there was little change in the edge electron density ne and electron temperature T e profiles. The density fluctuation amplitude was reduced approaching the transition, while a large increase of turbulence Reynolds stress and flow shear were simultaneously observed. The turbulence decorrelation rate was found to increase as the ion ∇B drift direction was moving towards the favourable configuration, but the flow shear also increased and exceeded the turbulence decorrelation rate. These measurements demonstrate an important correlation between turbulence and turbulence-driven flow and a lowering of PLH, provide insights into the underlying physics behind the hidden parameters and inform a more complete physics-based model of the L-H transition power threshold.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Investigation of the Suppression of Edge Localized Modes by Resonant Magnetic Perturbations in NSTX/NSTX-U (Final Technical Report)

The purpose of the grant was to investigate the suppression of edge localized modes by resonant magnetic perturbations in both NSTX and NSTX-U. Unfortunately, the project quickly exhausted all of the relevant experimental data from NSTX. It was originally intended to use new data from NSTX-U, when it restarted. However, the considerable delay in the restart made this impossible. Hence, with the permission of DOE, the grant was pivoted toward a more general examination of the MHD stability of tokamak plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

A novel measurement of marginal Alfven Eigenmode stability during high power auxiliary heating in JET

The interaction of Alfven Eigenmodes (AEs) and energetic particles is one of many important factors determining the success of future tokamaks. In JET, eight in-vessel antennas were installed to actively probe stable AEs with frequencies ranging 25-250 kHz and toroidal mode numbers |n| < 20. During the 2019-2020 deuterium campaign, almost 7500 resonances and their frequencies f0, net damping rates \gamma < 0, and toroidal mode numbers were measured in almost 800 plasma discharges. From a statistical analysis of this database, continuum and radiative damping are inferred to increase with edge safety factor, edge magnetic shear, and when including non-ideal effects. Both stable AE observations and their associated damping rates are found to decrease with |n|. Active antenna excitation is also found to be ineffective in H-mode as opposed to L-mode; this is likely due to the increased edge density gradient's effect on accessibility and ELM-related noise's impact on mode identification. A novel measurement is reported of a marginally stable, edge-localized Ellipticity-induced AE probed by the antennas during high-power auxiliary heating (ICRH and NBI) up to 25 MW. NOVA-K kinetic-MHD simulations show good agreement with experimental measurements of f0, \gamma, and n, indicating the dominance of continuum and electron Landau damping in this case. Similar experimental and computational studies are planned for the recent hydrogen and ongoing tritium campaigns, in preparation for the upcoming DT campaign.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

A novel measurement of marginal Alfvén eigenmode stability during high power auxiliary heating in JET

The interaction of Alfvén eigenmodes (AEs) and energetic particles is one of many important factors determining the success of future tokamaks. In JET, eight in-vessel antennas were installed to actively probe stable AEs with frequencies ranging 25–250 kHz and toroidal mode numbers |n| < 20. During the 2019–2020 deuterium campaign, almost 7500 resonances and their frequencies f 0 , net damping rates γ < 0, and toroidal mode numbers were measured in almost 800 plasma discharges. From a statistical analysis of this database, continuum and radiative damping are inferred to increase with edge safety factor, edge magnetic shear, and when including non-ideal effects. Both stable AE observations and their associated damping rates are found to decrease with |n|. Active antenna excitation is also found to be ineffective in H-mode as opposed to L-mode; this is likely due to the increased edge density gradient’s effect on accessibility and ELM-related noise’s impact on mode identification. A novel measurement is reported of a marginally stable, edge-localized ellipticity-induced AE probed by the antennas during high-power auxiliary heating (ion cyclotron resonance heating and neutral beam injection) up to 25 MW. NOVA-K kinetic-MHD simulations show good agreement with experimental measurements of f 0 , γ, and n, indicating the dominance of continuum and electron Landau damping in this case. Similar experimental and computational studies are planned for the recent hydrogen and ongoing tritium campaigns, in preparation for the upcoming DT campaign.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Vortex splitting in two-dimensional fluids and non-neutral electron plasmas with smooth vorticity profiles

Initially, elliptical, quasi-two-dimensional (2D) fluid vortices can split into multiple pieces if the aspect ratio is sufficiently large due to the growth and saturation of perturbations known as Love modes on the vortex edge. Presented here are experiments and numerical simulations, showing that the aspect ratio threshold for vortex splitting is significantly higher for vortices with realistic, smooth edges than that predicted by a simple “vortex patch” model, where the vorticity is treated as piecewise constant inside a deformable boundary. The experiments are conducted by exploiting the E × B drift dynamics of collisionless, pure electron plasmas in a Penning–Malmberg trap, which closely model 2D vortex dynamics due to an isomorphism between the Drift–Poisson equations describing the plasmas and the Euler equations describing ideal fluids. The simulations use a particle-in-cell method to model the evolution of a set of point vortices. The aspect ratio splitting threshold ranges up to about twice as large as the vortex patch prediction and depends on the edge vorticity gradient. This is thought to be due to spatial Landau damping, which decreases the vortex aspect ratio over time and, thus, stabilizes the Love modes. Near the threshold, asymmetric splitting events are observed in which one of the split products contains much less circulation than the other. These results are relevant to a wide range of quasi-2D fluid systems, including geophysical fluids, astrophysical disks, and drift-wave eddies in tokamak plasmas.

Physics↗

Internal measurements of electromagnetic geodesic acoustic mode (GAM) in EAST plasmas

Velocity, density, and magnetic fluctuations of the geodesic acoustic mode (GAM) have been measured using the Doppler backscattering system, Faraday-effect polarimeter-interferometer, and external pick-up coils in the Experimental Advanced Superconducting Tokamak. Simultaneous measurements of density and velocity fluctuations at the midplane and top of plasmas demonstrate that m = 1 density fluctuations are quantitatively balanced by the compression of perpendicular flow fluctuations. Furthermore, internal magnetic fluctuations associated with GAM have now been directly measured by laser-based Faraday-effect polarimetry for the first time. Line-averaged magnetic fluctuations (up to 16 Gauss, B̃¯R,GAMBT∼0.066%) are significantly larger than those extrapolated from edge coils (a few Gauss) and that magnetic fluctuations increase with β. The observed discrepancy between finite β theory and experimental data indicates the need for further theoretical investigations.

Physics↗

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↗