Confinement Time Research Articles

Simulation of an ohmic regime in the T-15MD tokamak based on the canonical profile transport model

mail: nrcki@nrcki.ru The canonical profiles transport model (CPTM), whose coefficients were determined from the T-10 tokamak database with a standard magnetic field BT = 2.3—2.5 T, has shown its robustness in ohmic regimes with a reduced magnetic field BT = 1.55—2.1 T. We used the CPTM for predictions of radial profiles and dependences of the electron and ion temperatures and the energy confinement time on the average plasma density for the T-15MD tokamak at the initial stage of its operation: the ohmic regime in a circular limiter configuration with BT = 1.0—2.0 T and plasma current Ip 1 MA.

A necessary condition for breakeven in dipole-confined plasmas

We have derived a necessary condition for the achievement of breakeven in axisymmetric plasmas with zero toroidal field and confined by a dipole magnetic field (B. Lehnert, Nature, vol. 181, 1958, p. 4605; A. Hasegawa, Comments Plasma Phys. Control. Fusion, vol. 11, no. 3, 1987). Excellent MHD stability, high values of $\beta$ (up to $26$ %) and good confinement properties awaken the interest of private investors after years of neglect due to lack of public funding and competing alternative lines of research like the tokamak. Starting from a requirement of self-consistency between the balances of momentum and energy in a dipole-confined, two-species plasma and assuming a Maxwellian distribution function for ions and electrons, we derive a necessary condition for breakeven. This condition is more stringent than the Lawson criterion because of the lack of a stabilizing toroidal field. For a given current flowing across the toroidal coil internal to the plasma, the crucial factor at stake is the ratio between the radius of the main toroidal coil and the radius of the vacuum chamber.

Full-f gyrokinetic simulations of hydrogen isotope mixing in tokamak plasmas

Hydrogen isotope mixing phenomena in tokamak plasmas are analyzed using global full-f gyrokinetic simulations. Model plasma parameters are chosen based on hydrogen isotope pellet experiments on JET, in which fast hydrogen isotope mixing in the timescale of the energy confinement time τE occurred after injecting deuterium (D) pellets into hydrogen (H) plasmas. Two numerical experiments are conducted using plasma profiles before and after the pellet injection. In both cases, turbulent fluctuations in the plasma core are characterized by ion temperature gradient driven turbulence. In the former case, the density profile of bulk H ions is kept in a quasi-steady state, and the particle confinement time of bulk H ions τH is an order of magnitude longer than τE. In the latter case, the density profiles of bulk H ions and pellet D ions show transient relaxation in the timescale of τE, and both τH and the timescale of particle pinch τD become comparable to τE, indicating the fast hydrogen isotope mixing. In the toroidal angular momentum balance, it is found that the enhanced ion particle transport leading to the fast hydrogen isotope mixing is driven by the toroidal field stress.

Thermonuclear performance variability near ignition at the National Ignition Facility

We describe our current understanding of the variability and degradation mechanisms observed through a series of five indirectly driven inertial fusion implosions fielded at the National Ignition Facility in the fall of 2021, four of which attempted to reproduce the first experiment to achieve Lawson's criterion for ignition with a thermonuclear yield of 1.35 MJ on August 8, 2021. A large number of absolutely calibrated (imaging, time-resolved, and spectrally resolved) x-ray and neutron diagnostics are fielded on the NIF along multiple lines of sight for each experiment. This allows for a reconstruction of the DT fuel and ablator mix injected into the hotspot around peak burn. We show that nuclear yield variations are well reproduced by numerical modeling when the measured low mode asymmetries and mix mass are included. Furthermore, these observed perturbations during burn are linked to small variations in laser delivery and capsule defects. Stringent specifications are then set to achieve robust ignition with the implosion design studied in this paper.

Effect of rotation on negative triangularity plasmas in DIII-D

Abstract At the DIII-D tokamak, plasmas with negative triangularity (NT) have been found to have similar energy confinement times as H-mode plasmas with positive triangularity. These NT plasmas also have a stable edge that inhibits H-mode, lacks dangerous edge transients, and has naturally low impurity confinement. For these reasons, NT plasmas have the potential to be a transformative reactor scenario. To determine the importance of strong co-current neutral beam torque for NT performance at DIII-D, scans of neutral beam torque at constant neutral beam power have been performed. These scans have shown that the energy confinement time for a basic, inductive NT scenario ( q 95 ≈ 2.8 ) decreases by 16% when torque is reduced to reactor levels. This reduction in confinement can be replicated with gyrofluid modeling that isolates the role of E × B shear in the change in transport. In a separate set of discharges with lower plasma current ( q 95 ≈ 4.2 ), no significant correlation between confinement time and rotation was found. Throughout this investigation, excellent MHD stability was observed, even with very low toroidal rotation. Further, low impurity confinement was found at multiple rotation levels. Throughout the campaign, rotation near the edge was counter-current, consistent with the measurement of counter-current intrinsic edge rotation, and these observations agree with NT results from the TCV tokamak.

Comment on "Ultracold plasma expansion in quadrupole magnetic field".

Bronin etal. [Phys. Rev. E 108, 045209 (2023)2470-004510.1103/PhysRevE.108.045209] recently reported molecular-dynamics simulations of ultracold neutral plasmas expanding in a quadrupole magnetic field. While the main results are in agreement with prior experimental measurements, we present data showing oscillations not captured in the simulations of Bronin etal. Plasmas formed using pulsed or continuous-wave ionization processes have similar confinement times.

Observation of tungsten emission spectra up to W46+ ions in the Large Helical Device and contribution to the study of high-Z impurity transport in fusion plasmas

Spectroscopic studies of emissions released from tungsten ions combined with a pellet injection technique have been conducted in Large Helical Device for contribution to the tungsten transport study in tungsten divertor fusion devices and for expansion of the experimental database of tungsten line emissions. The spectral intensities of W5+, W24+–W28+, W37+, W38+, W41+–W43+, W45+, and W46+ emission lines were measured simultaneously over a wide wavelength range from x-ray to visible. Time evolutions of the various tungsten line spectra indicate that the tungsten confinement time depends on the electron density of the plasma and is long in high density plasmas, on the order of seconds, and short in low density plasmas, on the order of sub-seconds. When the confinement time was long, the tungsten ions remained in the plasma until the end of the discharge, changing their dominant charge with the change in electron temperature. When the confinement time was short, the tungsten ions rapidly decreased in all charge states and disappeared. Space-resolved EUV and visible spectroscopy measurements have revealed that tungsten ions stayed in the core region of the plasma with changing their dominant charge state depending on the electron temperature in the discharges with the long confinement time. Detailed analysis of soft x-ray emission suggested that the confinement time increases with density and becomes saturated when the central electron density exceeds 2 × 1013 cm−3.

Simultaneous access to high normalized density, current, pressure, and confinement in strongly-shaped diverted negative triangularity plasmas

Strongly-shaped diverted negative triangularity (NT) plasmas in the DIII-D tokamak demonstrate simultaneous access to high normalized density, current, pressure, and confinement. NT plasmas are shown to exist across an expansive parameter space compatible with high fusion power production, revealing surprisingly good core stability properties that compare favorably to conventional positive triangularity plasmas in DIII-D. Non-dimensionalizing the key parameters, expanded operating spaces featuring edge safety factors below 3, normalized betas above 3, Greenwald density fractions above 1, and high-confinement mode (H-mode) confinement qualities above 1 are observed, even simultaneously, and all with a robustly stable edge free from deleterious edge-localized mode instabilities. Scaling of the confinement time with engineering parameters reveals at least a linear dependence on plasma current although with significant power degradation, both in excess of expected H-mode scalings. These results increase confidence that NT plasmas are a viable approach to realize fusion power and open directions for future detailed study.

Injection and confinement of positron bunches in a magnetic dipole trap.

We demonstrate the efficient injection of a pulsed positron beam into a magnetic dipole trap and investigate the ensuing particle dynamics in the inhomogeneous electric and magnetic fields. Bunches of ∼10^{5}e^{+} were transferred from a buffer-gas trap into the field of a permanent magnet using a lossless E×B drift technique. The Δt≈0.2µs pulses were short compared to the toroidal rotation period, τ_{d}≈16µs, and e^{+} confinement time, τ_{c}≈0.6s. The redistribution dynamics were studied by measuring the delayed γ-ray emission as the trap was emptied. This work extends the record for the number of low-energy positrons held in a dipole trap by two orders of magnitude and represents a significant advance toward the confinement of an electron-positron pair plasma.

Electron confinement investigation in low beta magnetic polywell configurations

This study uses numerical simulations to investigate the effects of magnetic field topology resulting from various polywell fusion setup configurations, including the cube configuration (6 coils), dodecahedron configuration (12 coils), double-layer configuration (14 coils) and disco configuration (26 coils). The results suggest that increased number of magnetic coils and magnitude of magnetic flux density through increased coil current leads to a longer electron confinement time. This is shown by the increased magnetic flux density and magnetic well width with increasing number of coils. In addition, each configuration is investigated to predict the capacity of electron confinement. Numerical electron injections are applied to each magnetic field topology to determine the decay behavior of electron numbers, from which the electron confinement time is calculated.

Retraction Note: Study of Effects of the Effective Edge Safety Factor on the Energy Confinement Time in IR-T1 Tokamak

Retraction Note: Study of Effects of the Effective Edge Safety Factor on the Energy Confinement Time in IR-T1 Tokamak

Plasma performance and operational space with an RMP-ELM suppressed edge

The operational space and global performance of plasmas with edge-localized modes (ELMs) suppressed by resonant magnetic perturbations (RMPs) are surveyed by comparing AUG, DIII-D, EAST, and KSTAR stationary operating points. RMP-ELM suppression is achieved over a range of plasma currents, toroidal fields, and RMP toroidal mode numbers. Consistent operational windows in edge safety factor are found across devices, while windows in plasma shaping parameters are distinct. Accessed pedestal parameters reveal a quantitatively similar pedestal-top density limit for RMP-ELM suppression in all devices of just over 3×1019 m−3. This is surprising given the wide variance of many engineering parameters and edge collisionalities, and poses a challenge to extrapolation of the regime. Wide ranges in input power, confinement time, and stored energy are observed, with the achieved triple product found to scale like the product of current, field, and radius. Observed energy confinement scaling with engineering parameters for RMP-ELM suppressed plasmas are presented and compared with expectations from established H and L-mode scalings, including treatment of uncertainty analysis. Different scaling exponents for individual engineering parameters are found as compared to the established scalings. However, extrapolation to next-step tokamaks ITER and SPARC find overall consistency within uncertainties with the established scalings, finding no obvious performance penalty when extrapolating from the assembled multi-device RMP-ELM suppressed database. Overall this work identifies common physics for RMP-ELM suppression and highlights the need to pursue this no-ELM regime at higher magnetic field and different plasma physical size.

Retraction Note: Experimental Study of Effects of the Internal Inductance on the Energy Confinement Time in IR-T1 Tokamak

Retraction Note: Experimental Study of Effects of the Internal Inductance on the Energy Confinement Time in IR-T1 Tokamak

Retraction Note: Measurement of Plasma Energy Confinement Time in Presence of Resonant Helical Field in IR-T1 Tokamak

Retraction Note: Measurement of Plasma Energy Confinement Time in Presence of Resonant Helical Field in IR-T1 Tokamak

Plasma and gas neutralisation of high energy H- and D- in an Argon plasma neutraliser

A beam driven plasma neutraliser (BDPN) has been proposed [1] as a relatively simple way to increase the efficiency of neutral beam injectors based on the acceleration and neutralisation of H- or D-. Initial calculations showed that sufficient levels of ionisation could be achieved with H2 or D2 as the initial gas target in the neutraliser if a plasma confinement time of the plasma in the neutraliser is >0.5 ms could be achieved. However, later calculations [2,3] show that the afore-mentioned results were very optimistic as the calculations did not take into account the recombination of molecular ions with electrons. Hence, it has been suggested [4] to use an argon (Ar) plasma as the molecular ion content in such a plasma will be negligible. The use of an Ar plasma neutraliser has been suggested previously [5–7], but few details have been given of the obtainable neutralisation efficiency, the necessary line density for that neutralisation efficiency, or of the calculations and cross sections used to deduce those quantities. This paper briefly discusses why Ar could be a good choice for a plasma neutraliser, then the reactions occurring between the H-/D- beam and the particles in the Ar plasma, and the available cross section data for those reactions. Subsequently, the differential equations for the change of the species in the beam as it traverses an Ar plasma are deduced and solved, and the solutions used to calculate the achievable neutralisation efficiency as a function of the line density in the neutraliser etc. Results are given of the neutralisation efficiency as a function of the degree of ionisation in the neutraliser, the required line density, and the species changes in the beam for accelerated H- beams with 100, 500 and 870 keV energies.

Improvement in ion confinement time with multigrid configuration in an inertial electrostatic confinement fusion device.

Improvement in the functionality of an inertial electrostatic confinement fusion (IECF) device has been investigated through kinetic simulation. Previously, we achieved a neutron generation rate of 10^{6} neutrons per second, but higher rates and better plasma confinement are necessary for broader applications. We compared a traditional single-grid IECF device with a triple-grid variant to evaluate the benefits of using multiple grids for ion confinement. Our computational models, using the 2D-3V xoopic code, suggest that the triple-grid device, with its optimized potentials, could significantly enhance ion confinement. The models show that the triple-grid design directs ion beams more effectively to the center, in contrast with the more scattered ion distribution in the single-grid design. This results in longer ion lifetimes in the triple-grid system due to its modified electrostatic fields. In the standard single-grid IECF device, the primary reasons for ion loss are chaotic ion trajectories and interactions with residual gases. By operating the triple-grid device under very low background gas pressure and with a focused field structure, we expect to achieve improved ion confinement.

Retraction Note to: TF Ripple Effects on Plasma Energy Confinement Time in IR-T1 Tokamak

Retraction Note to: TF Ripple Effects on Plasma Energy Confinement Time in IR-T1 Tokamak

Stellarators with enhanced tritium confinement and edge radiation control

A stellarator design is described with the purpose of achieving three goals: (1) enhance the confinement time of tritium. (2) Have a sufficient density of high-Z impurities to radiate the thermal power escaping from the core while having an extremely low impurity density in the core. (3) Maintain a large fraction of the plasma in a burning plasma state with an optimal tritium fraction. Some features of this design could be used in tokamaks. Although having three confinement zones is natural for stellarators, it is not for tokamaks.

What next: Further implosion space exploration on the path to NIF extended yield capability

We present quantitative motivations and assessments of various proposed and ongoing directions to further improving yields and target gain of igniting indirect-drive implosions at the National Ignition Facility (NIF). These include increasing compression and confinement time, improving hohlraum and ablator efficiency, and further increasing peak power and laser energy. 1D hydroscaled simulations, augmented by analytic 1D theory, have been used to project yield improvements for each of these implosion optimization tracks, normalized to the best current performing 4 MJ shot. At current NIF capabilities of 2.2 MJ, 450 TW, we project several paths could reach 15 MJ yield levels. We also expect several key implosion physics questions will be addressed in attempting to reach this yield level. These include demonstrating to what extent lower adiabat designs leading to higher compression will increase gain and efficiency, and whether we can reduce residual kinetic energy and ablator-fuel mix that is probably limiting the current burn-up fraction. For an envisaged NIF upgrade to EL = 3 MJ at fixed 450 TW peak power, scaling capsule size and fuel thicknesses faster than pure hydroscaling should allow for yields that could reach up to 60–80 MJ, depending on the efficiency gains realized in increasing deuterium-tritium fuel thickness, reducing hohlraum losses, and switching to lower Z ablators. The laser-plasma instability and beam transmission scaling in these larger hohlraums is shown to be favorable if the spot size is increased with hohlraum scale.

Assessing core ion thermal confinement in critical-gradient-optimized stellarators

We investigate the core confinement properties of two recently devised quasi-helically symmetric stellarator configurations, HSK and QSTK. Both have been optimized for large critical gradients of the ion temperature gradient mode, which is an important driver of turbulent transport in magnetic confinement fusion devices. To predict the resulting core plasma profiles, assuming a fixed edge temperature, we utilize an advanced theoretical framework based on the gyrokinetic codes GENE and GENE-3D, coupled to the transport code TANGO. Compared to the HSX stellarator, both HSK and QSTK achieve significantly higher core-to-edge temperature ratios, partly thanks to their smaller aspect ratios, with the other part due to more detailed shaping of the magnetic geometry achieved during optimization. The computed core confinement time, however, is less sensitive to core temperature than the fixed edge temperature, simply due to the disproportionate influence, the edge has on stored plasma energy. We, therefore, emphasize the possible benefits of further optimizing turbulence in the outer core region, and the need to include accurate modeling of confinement in the edge region in order to assess overall plasma performance of turbulence optimized stellarators.