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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Impurity radiation seeding of neoclassical tearing mode growth

Abstract The physics of neoclassical tearing modes (NTMs) is of great concern to tokamak plasma stability and performance, especially in the burning plasma regime. Whereas in many situations the seed events can be clearly identified, such as sawteeth and edge localized modes, the potential seeding mechanism of NTMs due to the resistive tearing instability driven by the impurity radiation cooling still needs more study. Recent NIMROD simulations have demonstrated that local impurity radiation cooling can drive the seed island growth and trigger the subsequent onset of NTM instability. The seed island is mainly driven by the local helical perturbation of the diamagnetic current induced by the perturbed pressure gradient as a result of the impurity radiative cooling on the rational surface. A heuristic closure for the neoclassical viscosity is adopted, and the seed island is further driven by the perturbed bootstrap current induced from the neoclassical electron viscous stress in the extended Ohm’s law. The growth rate of the NTM in simulations is found proportional to the electron neoclassical viscosity, and a theoretical neoclassical driving term is adopted to account for the nonlinear neoclassical island growth in the simulations.

Zeng, Shiyong (ORCID:000000017098818X)↗
DOE OSTI2025

HIPED: Machine learning framework for spherical tokamak pedestal prediction and optimization

We introduce a Machine Learning framework, HIPED (HeIght and width Predictor for Edge Dynamics), for predicting and optimizing pedestal and core performance in spherical tokamak plasmas. Trained on pedestal and core datasets from the third MAST-U campaign, HIPED provides accurate estimates of pedestal height and width. The results reveal notable differences compared with conventional aspect-ratio studies; for instance, a simple power-law relation between pedestal width and height has very low accuracy. Instead, additional parameters such as normalized plasma pressure, elongation, and Greenwald fraction significantly improve accuracy. HIPED can also be trained only on `control room parameters' to inform experimentalists of which controllable parameters to adjust for improving core-integrated performance. The framework further includes a multi-objective optimization scheme that helps guide experimental planning and optimization. We find Pareto-optimal discharges with respect to various features, including distance from edge-localized modes and normalized plasma pressure, track their parameter trajectories over time, and identify the control room parameters required for these Pareto-optimal discharges. This provides a framework for systematically optimizing core and edge performance according to different experimental priorities.

Parisi, Jason F. [Princeton Plasma Physics Laborat↗
DOE OSTI2020

Advanced grazing-emission X-ray fluorescence spectroscopy with small-angle neutron scattering for in-vivo surface composition and defect/morphology surface evolution in tokamak PMI

The dynamic and extreme conditions of thermonuclear fusion tokamak plasmas render material surfaces almost impossible to examine in real-time, surface-sensitive conditions. However, combining particle-probe techniques that are not charge-state dependent open the possibilities of in-vivo diagnosis of the plasma-material interface. The reconstituted surfaces of plasma-facing components ranging from first-wall to divertor-component regions in the device vary depending on the complex plasma transport characteristics of the device. For example, in ITER it is predicted that fluxes of the order 10 22 to 10 24 and incident charged-particle energies of 100’s to 1000’s of eV will reach the surface while in the divertor regions within the private-flux and near the inner and outer-strike points would vary from a few eV to 100’s of eV governed by the complex Chodura magnetic sheath dynamics. This means that the penetration range of most charged-particle fluxes would range between 1-100’s of nm (e.g. for fuel particles of D and T and He).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Unveiling the Significance of Correlations in K-Space and Configuration Space for Drift Wave Turbulence in Tokamaks

Turbulence and transport phenomena play a crucial role in the confinement and stability of tokamak plasmas. Turbulent fluctuations in certain physical quantities, such as density or temperature fluctuations, can have a wide range of spatial scales, and understanding their correlation length is important for predicting and controlling the behavior of the plasma. The correlation length in the radial direction is identified as the critical length in real space. The dynamics in real space are of significant interest because transport in configuration space is primarily focused on them. When investigating transport caused by the E×B drift, the correlation length in real space represents the size of E×B whirls. It was numerically discovered that in drift wave turbulence, this length is inversely proportional to the normalized mode number of the fastest growing mode relative to the drift frequency. Considerable time was required before a proper analytical derivation of this condition was accomplished. Therefore, a connection has been established between phenomena occurring in real space and those occurring in k-space. Although accompanied by a turbulent spectrum in k-space with a substantial width, transport in real space is uniquely determined by the correlation length, allowing for accurate transport calculations through the dynamics of a single mode. Naturally, the dynamics are subject to nonlinear effects, with resonance broadening in frequency being the most significant nonlinear effect. Thus, mode number space is once again involved. Resonance broadening leads to the detuning of waves from particles, permitting a fluid treatment. It should be emphasized that the consideration here involves the total electric field, including the induction part, which becomes particularly important at higher beta plasmas.

Weiland, Jan (ORCID:0000000253496025)↗
DOE OSTI2026

Experimental study of ECH pre-ionization on J-TEXT

An experimental study on electron cyclotron heating (ECH) pre-ionization has been conducted on J-TEXT in support of the joint experiment research for ITER plasma initiation. In this experiment, ECH power was injected to the vessel before the application of loop voltage ionize the neutral gas and form the initial plasma or so-called pre-plasma. The impact of several significant factors, such as magnetic field configuration, pre-fill gas pressure, ECH toroidal injection angle and ECH power on the evolution of pre-plasma are systematically studied, aiming to identify shared features, clarify their potential relationship and optimize the discharge parameters to generate a rather high pre-plasma density. By separating ECH power from the inductive start-up, the effect of pre-plasma on tokamak start-up can be observed. A dynamic magnetic configuration facilitates the transition of pre-plasma to tokamak plasma. To assess the influence of pre-plasma density on tokamak start-up, two kinds of magnetic field configurations are examined. While the effect of different pre-plasma densities on tokamak start-up is negligible, a significant difference is observed between pure ohmic start-up and start-up with pre-ionization. The studies presented here show evolutionary trends and threshold values needed to optimize ECH pre-ionization and also a feasible way to improve pre-plasma density and a configuration to stabilize pre-plasma for transition. Eventually, these results may contribute to multi-machine research and physics analysis that can assist ITER with its optimal preparation for first plasma operation.

ECH↗
DOE OSTI2025

Feasibility of main thermal ion heating by ICRF waves using a top launcher in a tokamak with deuterium–tritium plasmas

A scenario of ion cyclotron range of frequency (ICRF) wave injection from a top launcher is proposed as an efficient and direct heating method for thermal deuterium ions in deuterium–tritium tokamak plasmas. Positioned between the tritium cyclotron layer and ion–ion hybrid layer, the top launcher allows effective wave penetration to the ion–ion hybrid layer and enables significant power transfer to thermal deuterium. This is achieved through favorable wave polarization for fundamental cyclotron damping. There is a Doppler broadening around the cyclotron resonance and this overlaps with the ion–ion hybrid layer. Low toroidal mode numbers and ion temperature in the range of 5–20 keV are favorable for enhancing the main ion damping relative to electron damping. In contrast to the neutral beam injection, which penetration strongly depends on machine size and plasma density, the proposed ICRF-based direct ion heating scenario is shown to be scalable and applicable to both larger and smaller tokamak devices within practical constraints.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Investigation of the compatibility of pellet fueling with ELM-free H-mode plasmas in EAST tokamak

Abstract Experiments on pellet fueling have been carried out in edge localized mode (ELM)-free high-confinement mode (H-mode) plasmas with q 95 ∼ 6 in the EAST tokamak. Cryogenic deuterium pellets were injected into the ELM-free plasmas at a frequency of 10 or 5 Hz from ∼45 cm above the mid-plane on the low-field side. It is found that the ELM-free H-mode plasmas are still sustained even if both the edge and core plasma are impacted by the pellet injections (PIs). Several small ELMs would appear and the edge coherent mode accompanying the ELM-free phase fades or even disappears just after the pellet events, but the plasma would rapidly recover to the ELM-free state. Although the ELMing phase is very short, it may be an issue that still needs to be resolved in the future. Meanwhile, the 3/2 tearing mode often appearing in ELM-free discharges would be stabilized by the PIs, and the high-Z impurity concentration would be reduced during the PIs. All these results will be meaningful to International Thermonuclear Experimental Reactor and future fusion reactors.

Physics↗
DOE OSTI2021

Characteristics of Ultralow Aspect Ratio Toroidal Plasmas and Non-Solenoidal Startup and Edge Stability Studies at Near-Unity Aspect Ratio (Final Report)

This grant supported a long-term research and development program to explore the physics and technology of spherical tokamak (ST) plasma behavior at near-unity aspect ratio ( A ). The PEGASUS Toroidal Experiment is a university-scale ( R ) ~ 0.40 m, α ~ .35 m, I p ≤ 0.3 MA, B T ≤ 0.15 T) experiment that accesses A → 1, to provide large values of the normalized plasma current, I N = I p /α B T , or equivalently the toroidal field utilization factor I p / I TF . The research activities focused on: 1) developing the technique of Local Helicity Injection (LHI) as a viable means of initiating and growing an ST plasma without the use of a central solenoid; 2) exploring the equilibrium and stability properties of spherical tokamak plasmas at extremely low aspect ratio; and 3) obtaining new insights in the properties of the enhanced confinement H-mode regime at A ~ 1. The initial (Phase I) facility employed a unique high-stress central solenoid Ohmic for current drive and heating with simple magnet current waveforms. A subsequent major upgrade (Phase II) offered more flexible control of the plasma evolution along with a powerful LHI system for plasma initiation, heating and sustainment. The LHI technique consists of injection of high-density current streams into vacuum toroidal and vertical magnetic fields. These streams become unstable and support magnetic reconnection to allow the system to relax into a tokamak-like magnetic configuration. This program focused on: 1) development of electron current sources that provide sufficiently high injection currents in the challenging edge plasma region; 2) validation of theoretical limits to achievable I p arising from magnetic helicity and energy conservation plus absolute limits arising from relaxation to a minimal energy states; 3) tests of a power balance model to describe the evolution of the plasma current I p (t); and 4) study of magnetic fluctuations and their role in the I p evolution. This culminated in the production of toroidal currents of I p ≥ 0.2 MA with only ~8 kA of injected current. Operation with Ohmic induction alone in Phase I created plasmas with soft limits on the achievable I p and field utilization ( I p / I TF ~ 1) due to the onset of large low-order (2/1 or 3/2) internal resonant tearing modes. With the expanded capabilities in Phase II operation, methods employed to mitigate those modes and achieve I p / I TF > 1 were: 1) I TF ramp-downs; 2) non-solenoidal startup via LHI; and 3) extremely low I TF Ohmic operations assisted by LHI current injector pre-ionization. Extensive use of LHI-driven startup and drive for non-solenoidal operation produced favorable current profiles to stably operate tokamak plasmas with unprecedentedly high I N = I p / aB T ~ 15 and I p / I TF > 2. With strong reconnection-driven anomalous ion heating, this in turn provided access to tokamak plasmas with B t up to 100%, a world record 2 times higher than that previously achieved in advanced or spherical tokamaks. This high B t regime had a large fraction (~50%) of the plasma volume in a region of an absolute minimum- B configuration. The MHD stability limit in this regime was the ideal external kink mode without a stabilizing wall. These results provided the first demonstration of access to tokamak plasmas with very high B t , high i N , high elongation κ, and low internal inductance ℓ i , and as such served as a proof-ofprinciple for access to this long-sought ST regime. The naturally low B T needed for stability at A ≤ 1.2 reduces the L-H mode power threshold, P LH , so that H-mode plasmas were achieved with Ohmic heating alone. The power threshold for the transition to the H-mode regime was found to be 10× the accepted values from the ITER international scaling, confirming initial trends indicated by experiments on MAST and NSTX, and suggesting that the physics behind this transition is not yet fully understood. The density dependence of the transition and the observed insensitivity to whether the plasma is bounded by a limiter or magnetic divertor are consistent with the recent FM 3 model. Multiple unstable toroidal magnetic modes were measured during ELM (Edge Localized Mode) crashes, with generally lower toroidal mode numbers than seen at higher A , which was attributed to the larger peeling mode instability drive at low A . First-ever edge current profile measurements through an ELM crash showed current-carrying filaments were generated, as seen in ELM simulations and in electromagnetic blob transport theory.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

A method for automated modeling of plasma shape control response in tokamaks

Here, a simple method commonly used in modeling of tokamak conductor and plasma response for plasma control has been extended to allow more general modeling of tokamak PF conductor circuits. This extension combines a previously existing method for integrating plasma response and tokamak circuits with a more capable circuit modeling technique derived from the theory that underlies present-day automatic circuit simulation codes. This combined method can incorporate any combination of inductors (e.g., coils), resistors, capacitors, voltage sources, and current sources, while retaining the same model structure and ease of interpretation as the simple method. For many tokamaks, use of this extended method would not be the preferred approach, but for certain devices or for specialized models of a device, it is useful and occasionally essential. For example, the method has enabled fast and automated modeling of the complicated and daily-changing circuit topology used in the DIII-D tokamak. This automated modeling is essential to the support of controller design for the frequently changing requirements for DIII-D plasmas. It is likely to also find application in the workflows for new device designs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Scattering of lower hybrid radio frequency waves by cylindrical turbulent structures in the plasma edge in tokamaks

Lower hybrid (LH) radio frequency (RF) waves are used in fusion tokamak devices to generate non-inductively toroidal currents. LH waves are effective in imparting toroidal momentum to electrons in the core of the confined plasma. Since the LH waves are generated by wave guide structures near the wall of a tokamak, the waves have to propagate through a turbulent plasma in the edge region before coupling to the core. It is especially important to quantify the effect of this plasma region on the propagation characteristics of the LH waves, since fusion reactors like ITER will have an extended edge region. Structures in the plasma edge like filaments and blobs which have highly varying density fluctuations compared to the background density, span radial spatial scales that are comparable to the LH wavelength. We study the scattering of the LH waves by the filamentary structures using Maxwell's equations, in which the plasma permittivity is given by the cold plasma dispersion tensor. The collisional absorption of LH waves in the edge region is included by a modification to the elements of the dispersion tensor. The filaments are assumed to be cylindrical with the axis predominantly aligned along the direction of the toroidal magnetic field. Our studies are both analytical and numerical [1,2] and show that these structures can lead to reflection, refraction, diffraction, and side-scattering of both an incident LH plane wave and a Gaussian beam. We will present a variety of different density variations and collisional absorption rates in plasmas with filamentary structures of varying sizes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Staged cooling of a fusion-grade plasma in a tokamak thermal quench

Abstract In tokamak disruptions where the magnetic connection length becomes comparable to or even shorter than the plasma mean-free-path, parallel transport can dominate the energy loss and the thermal quench of the core plasma goes through four phases (stages) that have distinct temperature ranges and durations. The main temperature drop occurs while the core plasma remains nearly collisionless, with the parallel electron temperature T e ∥ dropping in time t as T e ∥ ∝ t − 2 and a cooling time that scales with the ion sound wave transit time over the length of the open magnetic field line. These surprising physics scalings are the result of effective suppression of parallel electron thermal conduction in an otherwise bounded, quasineutral, and collisionless plasma, which is different from what are known to date on electron thermal conduction along the magnetic field in a nearly collisionless and quasineural plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Improved Particle Confinement with Resonant Magnetic Perturbations in DIII-D Tokamak H-Mode Plasmas

Experiments on the DIII-D tokamak have identi fied a novel regime in which applied resonant magnetic perturbations (RMPs) increase the particle con finement and overall performance. This work details a robust range of counter-current rotation over which RMPs cause this density pump-in effect for high con finement (H-mode) plasmas. The pump-in is shown to be caused by a reduction of the turbulent transport and to be correlated with a change in the sign of the induced neoclassical transport. Furthermore, this novel reversal of the RMP induced transport has the potential to signifi cantly improve reactor relevant, three-dimensional magnetic con finement scenarios.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Tailoring tokamak error fields to control plasma instabilities and transport

A tokamak relies on the axisymmetric magnetic fields to confine fusion plasmas and aims to deliver sustainable and clean energy. However, misalignments arise inevitably in the tokamak construction, leading to small asymmetries in the magnetic field known as error fields (EFs). The EFs have been a major concern in the tokamak approaches because small EFs, even less than 0.1%, can drive a plasma disruption. Meanwhile, the EFs in the tokamak can be favorably used for controlling plasma instabilities, such as edge-localized modes (ELMs). Here we show an optimization that tailors the EFs to maintain an edge 3D response for ELM control with a minimized core 3D response to avoid plasma disruption and unnecessary confinement degradation. We design and demonstrate such an edge-localized 3D response in the KSTAR facility, benefiting from its unique flexibility to change many degrees of freedom in the 3D coil space for the various fusion plasma regimes. This favorable control of the tokamak EF represents a notable advance for designing intrinsically 3D tokamaks to optimize stability and confinement for next-step fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Parallel Algebraic Multigrid for Fusion and Higher-Order PDEs

Multigrid methods play a key role in large-scale scientific simulation because they are among the fastest and most scalable approaches for solving the underlying sparse linear systems of equations that arise from a wide array of Partial Differential Equation (PDE) discretizations. Algebraic multigrid (AMG) is a special type of multigrid method that depends only on the description of the linear system, giving it better portability and broader applicability than geometric multigrid, as it requires no explicit knowledge of the problem geometry. Even though these methods are widely used today, there are still applications where further development is needed. In this report, we focus on PDEs with higher-order terms (e.g., fourth order), concentrating on a PDE that arises in tokamak edge plasma simulations (a tokamak is a machine that confines a plasma using magnetic fields and is believed to be the leading plasma confinement concept for future fusion power plants). General multigrid relaxes a linear system on coarser grids and reverses this process with interpolation, but standard AMG methods struggle with the aforementioned higher-order PDEs. We investigate cyclic coarsening and interpolation heuristics, as well as new iterative approximation methods of refining the solution at each grid to improve the existing multigrid approach. To this end, we ensure that these techniques are transferable to a parallelized setting with LLNL’s supercomputers.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2026

Verification of global gyrokinetic simulation of low frequency mode excited by thermal plasma in spherical tokamak

A low frequency Alfven eigenmode in the ST40 spherical tokamak (ST) is found to be resonantly excited by thermal ions from global gyrokinetic GTC simulations. The analyses of phase-space resonances and wave-particle energy exchanges show that the Beta-induced Alfvén-Acoustic Eigenmode (BAAE) is primarily driven unstable by thermal ions, but is stabilized by energetic particles (EPs). When all kinetic effects are artificially suppressed, GTC simulations in the ideal magnetohydrodynamic (MHD) limit confirm the existence of the MHD eigenmode with a frequency in the BAAEs gap, as previously found by the ideal MHD code NOVA. The kinetic effects of thermal ions induce a frequency downshift compared to the stable BAAE in the MHD simulations. These results support the verification for global gyrokinetic simulations of the BAAE in the ST. However, the simulations do not explain the experimental observations of the BAAE in the ST40, possibly owing to the limitation of the collisionless gyrokinetic Monte Carlo model for the EPs.

Alfvén eigenmode↗
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↗