Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “plasma core”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Core plasma fueling by fast inward particle transport after hydrogen pellet injection in Wendelstein 7-X

A large database of more than 1000 individual cryogenic hydrogen pellets injected into Wendelstein 7-X for plasma fueling was analyzed to improve the understanding of the three phases of the process: the ablation, deposition and transport of the pellet material. Kilohertz-sampled electron density and temperature measurements revealed a more complex drift behavior than predicted by numerical code simulation. It could be explained by the poloidal plasma E r x B- drift rotation, which plays a significant role in stellarators, but was not previously considered in pellet injection codes like HPI2. The drift results in a fast poloidal rotation of the pellet material around the plasma core, leading to an almost homogeneous deposition over the involved flux surfaces regardless of magnetic high and low field side injection geometry. Additionally, a novel fast inward directed transport mechanism (‘FIT-effect’) was observed. The effect occurs on timescales of tens of milliseconds and cannot be explained by neoclassical transport or diffusion. It might be linked to the turbulence pinch recently found in Wendelstein 7-X. When the FIT-effect occurs, the pellet particles are rapidly transferred from the deposition flux surfaces to the plasma core, causing the plasma density profile to peak, which is beneficial for confinement in Wendelstein 7-X. The large pellet injection database was statistical analyzed with regard to pellet and plasma parameters, which delivered some starting points towards developing an understanding of the physics behind the FIT-effect. The results indicate, that plasma core fueling via pellet injection is largely independent of the injection geometry in stellarators under certain conditions, reducing the technical complexity of the injection system.

Wendelstein 7-X↗
DOE OSTI2026

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 ↗
DOE OSTI2021

Scoping study of detecting high harmonic fast waves in NSTX-U hot core plasma directly using beam emission spectroscopy

In this paper, we used radio frequency (RF) full wave simulation to explore the feasibility of generating a density beat oscillation formed by high harmonic fast waves (HHFWs) in the National Spherical Tokamak Experiment-Upgrade (NSTX-U) and of measuring it numerically with 2D beam emission spectroscopy (BES). This oscillation is of interest as it can be used to determine the injected RF fields. We began by computing HHFW fields in NSTX-U under a variety of experimental conditions and antenna configurations with Petra-M, a 3D RF wavefield solver. These fields were then used to evaluate the amplitude of δne,beat/ne0 in the volume of measurement, which we predict to be between 5 × 10−10 and 10−9, although it is highly dependent on plasma parameters. Trends in magnetic field, core temperature, core density, and antenna phasing are presented, and the parameters most suitable to this measurement are identified. Finally, we developed a synthetic BES diagnostic to compare the experimental BES signal against.

Rutherford, Grant (ORCID:0000000163620037)↗
DOE OSTI2020

Predictions of core plasma performance for the SPARC tokamak

SPARC is designed to be a high-field, medium-size tokamak aimed at achieving net energy gain with ion cyclotron range-of-frequencies (ICRF) as its primary auxiliary heating mechanism. Empirical predictions with conservative physics indicate that SPARC baseline plasmas would reach $\textit{Q}$ ≈ 11, which is well above its mission objective of $\textit{Q}$ > 2. To build confidence that SPARC will be successful, physics-based integrated modelling has also been performed. The TRANSP code coupled with the theory-based trapped gyro-Landau fluid (TGLF) turbulence model and EPED predictions for pedestal stability find that $\textit{Q}$ ≈ 9 is attainable in standard H-mode operation and confirms $\textit{Q}$ > 2 operation is feasible even with adverse assumptions. In this analysis, ion cyclotron waves are simulated with the full wave TORIC code and alpha heating is modelled with the Monte–Carlo fast ion NUBEAM module. Detailed analysis of expected turbulence regimes with linear and nonlinear CGYRO simulations is also presented, demonstrating that profile predictions with the TGLF reduced model are in reasonable agreement.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Milestone in predicting core plasma turbulence: successful multi-channel validation of the gyrokinetic code GENE

On the basis of several recent breakthroughs in fusion research, many activities have been launched around the world to develop fusion power plants on the fastest possible time scale. In this context, high-fidelity simulations of the plasma behavior on large supercomputers provide one of the main pathways to accelerating progress by guiding crucial design decisions. When it comes to determining the energy confinement time of a magnetic confinement fusion device, which is a key quantity of interest, gyrokinetic turbulence simulations are considered the approach of choice – but the question, whether they are really able to reliably predict the plasma behavior is still open. The present study addresses this important issue by means of careful comparisons between state-of-the-art gyrokinetic turbulence simulations with the GENE code and experimental observations in the ASDEX Upgrade tokamak for an unprecedented number of simultaneous plasma observables.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Role of microtearing mode in DIII-D and future high- β p core plasmas

This paper provides modeling evidence for a new variant of microturbulence instability that regulates the electron thermal fluxes in the transport barrier region of DIII-D "high β p " discharges. This scenario is being investigated as a possible scenario for future steady-state reactors, and thus it is essential to understand the physics processes which govern its performance in order to confidently extrapolate to future devices. Additional analysis examining the extrapolation of this scenario to ITER shows that while the new microtearing mode identified here is unlikely to be unstable in that regime (due to increased temperatures and reduced collisionality), the large shift in magnetic axis (Shafranov shift) that arises in these plasmas may still favor stabilization of common ballooning instabilities which peak on the low field side, but destabilize electrostatic slab modes on the high field side.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Predictions of core plasma performance for the SPARC tokamak

SPARC is designed to be a high-field, medium-size tokamak aimed at achieving net energy gain with Ion Cyclotron Range-of-Frequencies (ICRF) as its primary auxiliary heating mechanism. Empirical predictions with conservative physics indicate that SPARC baseline plasmas would reach Q~11, well above its mission objective of Q>2. To build confidence that SPARC will be successful, physics-based integrated modeling has also been performed. The TRANSP code coupled with the theory-based TGLF turbulence model and EPED predictions for pedestal stability find that Q~9 is attainable in standard H-mode operation and confirms Q>2 operation is feasible even with adverse assumptions. In this analysis, ion cyclotron waves are simulated with the full wave TORIC code and alpha heating is modeled with the Monte-Carlo fast ion NUBEAM module. Detailed analysis of expected turbulence regimes with linear and nonlinear CGYRO simulations is also presented, demonstrating that profile predictions with the TGLF reduced model are in reasonable agreement.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

From MAST-U to STEP: Power exhaust control challenges and opportunities

This work explores the challenges and opportunities for power exhaust control in STEP, informed by dedicated MAST-U experiments. The STEP system primarily relies on hydrogenic and argon gas injection into the divertors, compensating for transients originating in the plasma core. Fast transients (>10 Hz) that evolve too quickly for actuators to compensate must be absorbed passively; the enhanced transient buffering of the foreseen long-legged divertor provides a clear advantage over conventional geometries, as demonstrated in the MAST-U Super-X divertor. Experiments further indicate that STEP’s tight divertor baffling enables near-independent control of the upper and lower divertors. The most challenging transients are expected from power-sharing fluctuations, as MAST-U experiments observed extremely fast dynamics. Fluctuations from core pellet fuelling remain relatively benign due to the small pellet size according to simulations. Exhaust control in a reactor like STEP requires an integrated approach befitting its machine-critical nature. A predictive control element, integrated with core plasma control, allows pre-emptive preparation of the divertor for incoming transients. An observer-based diagnostic approach is foreseen to monitor the divertor in the challenging reactor environment, supported by an extended diagnostic set in the non-nuclear phase to validate the required dynamic models.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

First result of boronization assisted by the ICWC on EAST with full metal wall

Boron (B), a low-Z (atomic number) material, has been widely utilized in wall conditioning to improve plasma performance in fusion devices. In 2023, boronization was successfully conducted on EAST featuring an ITER-like tungsten divertor and fully metallic first wall. The process employed predischarge coating with carborane (C 2 B 10 H 12 ) as the working material, assisted by ion cyclotron wall conditioning (ICWC). After one time 12 g boronization, it was found the thickness of B film was approximately 120 nm. Post-boronization observations indicated that substantial hydrogen (H) release during initial plasma discharges compared with the consumed W/B wall, attributed to H co-deposition during the ICWC-boronization processing, which led to uncontrollable divertor neutral pressure and plasma density. The H/(H + D) ratio demonstrated a gradual reduction from ∼85% to 30% over more than 1850 s of deuterium plasma, with a cumulative injected energy of 2325 MJ. The B coating significantly enhanced the stored energy in plasma and improved confinement performance. The stored energy in plasma showed an increase of about 20%, primarily due to a reduction in impurity radiation, including oxygen (O) and heavy impurities such as tungsten (W), iron (Fe), and copper (Cu). The effective ion charge (Z eff ) decreased from 2.3 to 2.0. Following ICWC-boronization, the line-integrated radiation profile decreased by nearly 35% in the plasma core, plasma density and electron temperature exhibited an increase of ∼7% and 12% due to enhanced wall fueling and reduced impurity radiation. The lifetime of boronization, as evaluated by the line emissions from boron and other impurity radiation, was about 1700 s of deuterium plasma, with a cumulative injected energy of 2125 MJ on EAST. These findings provide significant insights for evaluating ICWC-boronization applicability in ITER with full W wall structure.

EAST↗
DOE OSTI2023

Core-Pedestal Plasma Configurations in Advanced Tokamaks

Here, several configurations for the core and pedestal plasma are examined for a predefined tokamak design by implementing multiple heating/current drive (H/CD) sources to achieve an optimum configuration of high fusion power in a noninductive operation while maintaining an ideally magnetohydrodynamic (MHD) stable core plasma using the IPS-FASTRAN framework. IPS-FASTRAN is a component-based lightweight coupled simulation framework that is used to simulate magnetically confined plasma by integrating a set of high-fidelity codes to construct the plasma equilibrium (EFIT, TOQ, and CHEASE), calculate the turbulent heat and particle transport fluxes (TGLF), model various H/CD systems (TORIC, TORAY, GENRAY, and NUBEAM), model the pedestal pressure and width (EPED), and estimate the ideal MHD stability (DCON). The TGLF core transport model and EPED pedestal model are used to self-consistently predict plasma profiles consistent with ideal MHD stability and H/CD (and bootstrap) current sources. In order to evaluate the achievable and sustainable plasma beta, varying configurations are produced ranging from the no-wall stability to with-wall stability regimes, simultaneously subject to the self-consistent TGLF, EPED, and H/CD source profile predictions that optimize configuration performance. The pedestal density, plasma current, and total injected power are scanned to explore their impact on the target plasma configuration, fusion power, and confinement quality. A set of fully noninductive scenarios are achieved by employing ion-cyclotron, neutral beam injection, helicon, and lower-hybrid H/CDs to provide a broad profile for the total current drive in the core region for a predefined tokamak design. These noninductive scenarios are characterized by high fusion gain (Q ~ 4) and power (P fus ~ 600 MW), optimum confinement quality (H 98 ~ 1.1), and high bootstrap current fraction (f BS ~ 0.7) for Greenwald fraction below unity. The broad current profile configurations identified are stable to low-n kink modes either because the normalized pressure β N is below the no-wall limit or a wall is present.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Self-consistent modeling of tokamak edge plasma transport with lithium sources

Magnetic confinement fusion devices require effective heat and particle exhaust solutions on the divertor plates to operate sustainably, especially under reactor-relevant conditions. Liquid lithium divertors have been proposed to address two major challenges: control of excessive heat flux to plasma-facing components through vapor shielding and minimization of core plasma contamination from impurities. The National Spherical Torus Experiment-Upgrade (NSTX-U) will explore lithium as a divertor material due to its potential to meet both objectives. We present a self-consistent coupling framework between the plasma boundary transport code UEDGE and the lithium wall transport code Wall–Li to evaluate the feasibility and operational limits of lithium-based divertors. The model aims to optimize lithium sourcing levels to prevent core plasma contamination via fuel dilution while ensuring divertor protection through vapor shielding. This integrated framework, applicable to any tokamak with lithium sources, dynamically adjusts lithium sourcing based on local plasma conditions and surface temperature. The coupled model is tested using NSTX-like geometry and plasma conditions to assess its performance and reliability. Wall–Li calculates lithium fluxes from plasma-facing components, incorporating physical sputtering, thermally enhanced sputtering, and evaporation driven by surface temperature and ion flux. These fluxes are reintroduced into UEDGE as neutral lithium atoms, enabling simulation of their transport and distribution within the plasma. UEDGE computes plasma and neutral transport, surface heat flux, and iteratively feeds this information back to Wall–Li. A small time step is employed to ensure numerical stability and convergence, enabling accurate simulations over typical tokamak discharge durations. This integrated modeling approach provides a robust tool for identifying operational regimes that balance effective lithium sourcing with minimal core plasma contamination, offering critical insights for optimizing lithium-based divertor systems in current and future fusion devices.

Magnetic confinement fusion↗
DOE OSTI2022

Modelling the Alfvén eigenmode induced fast-ion flow measured by an imaging neutral particle analyzer

An imaging neutral particle analyzer (INPA) provides energy and radially resolved measurements of the confined fast-ion population ranging from the high-field side to the edge on the midplane of the DIII-D tokamak. In recent experiments, it was used to diagnose fast-ion flow in the INPA-interrogated phase-space driven by multiple, marginally unstable Alfvén eigenmodes (AEs). Here, the key features of this measured fast-ion flow are: (I) a fast-ion flow from q min and the injection energy (81 keV) towards lower energies and plasma periphery.(II) A flow from the same location towards higher energies and the plasma core, (III) a phase-space 'hole' at the injected energy and plasma core and (IV) a pile-up at the plasma core at lower energies (~60 keV). Ad hoc energetic particle diffusivity modelling of TRANSP significantly deviates from the observation. Comparably, a reduced modelling, i.e. a combination of NOVA-K and ASCOT5 code with the measured mode structure and amplitude, generally reproduce some key features of the observed phase-space flow, but largely failed to interpret fast ion depletion near the plasma axis. At last, self-consistent, first-principle multi-phase hybrid simulations that include realistic neutral beam injection and collisions are able to reproduce most features of the time-resolved phase-space flow. During consecutive hybrid phases, an RSAE consistent with the experiment grows and saturates, redistributing the injected fast ions. The resulting synthetic INPA images are in good agreement with the measurement near the injection energy. The simulations track the fast-ion redistribution within the INPA range, confirming that the measured fast-ion flow follows streamlines defined by the intersection of phase-space surfaces of constant magnetic moment μ and constant E' = nE + ωP φ , where n and ω are the instability toroidal mode number and frequency, and E and P φ the ion energy and toroidal canonical momentum. Nonperturbative effects are required to reproduce the depletion of fast ions near the magnetic axis at the injection energy.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Observation and characterisation of trapped electron modes in Wendelstein 7-X

In the past, quasi coherent (QC) modes were reported for nearly all tokamaks. The general definition describes modes as QC when the magnitude squared coherence is in the range of 0.3 to 0.6. QC modes are observed in the plasma core as well as in the plasma edge and can have quite different physical origins. The one in the core are observed in plasmas with low collisionality, where the electron temperature exceeds the ion temperature in the plasma core. This is the case for electron cyclotron heating in general. The origin of these modes are electrons trapped within a magnetic mirror, as reported in the past from various fusion devices. The so-called trapped-electron modes (TEMs) belong to drift wave instabilities and can be destabilized by electron-temperature gradients in the plasma core. From the diagnostic point of view, QC modes appear as fluctuations in electron density and temperature. Therefore, the microwave reflectometer is very well suited to monitor these modes. This paper describes experiments, conducted at the Wendelstein 7-X stellarator (W7-X), which aim at detecting QC modes at low wave numbers. A poloidal correlation reflectometer installed at W7-X, is able to measure low wave numbers ($k_\perp\unicode{x2A7D}{3.5}\,\textrm{cm}^{-1}$). For medium line-averaged densities ($\int n_\textrm{e}\unicode{x2A7D}{6 \times10^{19}}\,\textrm{m}^{-2}$) the plasma core is accessible for this diagnostic. For different magnetic configurations and plasma parameters, broad QC structures are observed in the coherence spectra. From the analysis of the rotation and the poloidal structure, these QC modes show the properties of electron-temperature-gradient driven TEMs. A linear relation between the mode velocity and the rotation frequency is found. The relation is uniform and confirms the nature of QC-mode observation as TEM in tokamaks, too.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Achievements of actively controlled divertor detachment compatible with sustained high confinement core in DIII-D and EAST

The compatibility of efficient divertor detachment with high-performance core plasma is vital to the development of magnetically controlled fusion energy. The joint research on the EAST and DIII-D tokamaks demonstrates successful integration of divertor detachment with excellent core plasma confinement quality, a milestone towards solving the critical plasma–wall-interaction (PWI) issue and core-edge integration for ITER and future reactors. In EAST, actively controlled partial detachment with T et,div ~ 5 eV around the strike point and H 98 > 1 in different H-mode scenarios including the high β P H-mode scenario have been achieved with ITER-like tungsten divertor, by optimizing the detachment access condition and performing detailed experiments for core-edge integration. For active long-pulse detachment feedback control, a 30 s H-mode operation with detachment-control duration being 25 s has been successfully achieved in EAST. DIII-D has achieved actively controlled fully detached divertor with low plasma electron temperature (T et,div ≤ 5 eV across the entire divertor target) and low particle flux (degree of detachment, DoD > 3), simultaneously with very high core performance (β N ~3, β P > 2 and H 98 ~ 1.5) in the high βP scenario being developed for ITER and future reactors. The high-β P high confinement scenario is characterized by an internal transport barrier (ITB) at large radius and a weak edge transport barrier (ETB, or pedestal), which are synergistically self-organized. Both the high-β P scenario and impurity seeding facilitate divertor detachment. The detachment access leads to the reduction of ETB, which facilitates the development of an even stronger ITB at large radius in the high β P scenario. Thus, this strong large radius ITB enables the core confinement improvement during detachment. In conclusion, these significant joint DIII-D and EAST advances on the compatibility of high confinement core and detached divertor show a great potential for achieving a high-performance core plasma suitable for long-pulse operation of fusion reactors with controllable steady-state PWIs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Characterizing core and edge turbulence regimes with fluctuation imaging diagnostics in Wendelstein 7-X

Two density fluctuation imaging systems, phase contrast imaging (PCI) and gas puff imaging (GPI) measure spatially resolved density fluctuations with high time resolution throughout the core plasma (PCI) and in the scrape-off layer (GPI) of the Wendelstein 7-X (W7-X) stellarator. Both systems combined give a comprehensive overview of overall fluctuation levels, spectral properties such as their distribution in frequency and wavenumber space as well as their spatial distribution. These tools are used to assess changes in density turbulence in three representative discharges that transition into stable divertor detachment by different strategies (impurity seeding, density ramping and power starvation). Several general trends are identified when the radiated power fraction is systematically increased: In the plasma edge, the line emission observed by GPI shifts radially inward with a drop in electron temperature, and normalized intensity fluctuation profiles follow this inward shift. Skewness and kurtosis of these edge fluctuations are reduced, indicating a reduction of large intermittent transport events, and poloidal phase velocities decrease in magnitude. These observations are consistent with a reduced power input into the plasma edge and a general reduction of turbulent activity. Core density fluctuation levels remain nearly constant in the impurity seeding scenario, indicating that detachment does not significantly impact turbulence there. However, a strong reduction in the dominant outboard fluctuation phase velocity is observed that deviates from the previous interpretation of neoclassical radial electric field changes, showing that the core plasma is not completely unaffected. In the density ramp and power starvation scenarios, undesirable and irregular large-scale events arise clearly in both diagnostic systems as the radiative fraction is increased. Impurity seeding therefore seems to be a promising strategy on W7-X to achieve detachment without significantly altering core turbulence, especially when targeting a specific operating point in core density and heating power.

Wendelstein 7-X↗
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↗