Concerning a new theory of Type I comet tails
Enhanced density plasma confinement hypothesis for formation of Type I comet tails following interaction with solar wind
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Results for “PLASMA CONFINEMENT”
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.
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The alpha channeling effect
Alpha particles born through fusion reactions in a tokamak reactor tend to slow down on electrons, but that could take up to hundreds of milliseconds. Before that happens, the energy in these alpha particles can destabilize on collisionless timescales toroidal Alfven modes and other waves, in a way deleterious to energy confinement. However, it has been speculated that this energy might be instead be channeled into useful energy, so as to heat fuel ions or to drive current. Such a channeling needs to be catalyzed by waves Waves can produce diffusion in energy of the alpha particles in a way that is strictly coupled to diffusion in space. If these diffusion paths in energy-position space point from high energy in the center to low energy on the periphery, then alpha particles will be cooled while forced to the periphery. The energy from the alpha particles is absorbed by the wave. The amplified wave can then heat ions or drive current. This process or paradigm for extracting alpha particle energy collisionlessly has been called alpha channeling. While the effect is speculative, the upside potential for economical fusion is immense. The paradigm also operates more generally in other contexts of magnetically confined plasma.
71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
Resonance in Fast-Wave Amplitude in the Periphery of Cylindrical Plasmas and Application to Edge Losses of Wave Heating Power in Tokamaks
Heating magnetically confined plasmas using waves in the ion-cyclotron range of frequencies typically requires coupling these waves over a steep density gradient. This process has produced an unexpected and deleterious phenomenon on the National Spherical Torus eXperiment (NSTX): a prompt loss of wave power along magnetic field lines in front of the antenna to the divertor. Understanding this loss may be key to achieving effective heating and expanding the operational space of NSTX-Upgrade. Here, we propose that a new type of mode, which conducts a significant fraction of the total wave power in the low-density peripheral plasma, is driving these losses. We demonstrate the existence of such modes, which are distinct from surface modes and coaxial modes, in a cylindrical cold-plasma model when a half wavelength structure fits into the region outside the core plasma. The latter condition generalizes the previous hypothesis regarding the occurence of the edge losses and may explain why full-wave simulations predict these losses in some cases but not others. If valid, this condition implies that outer gap control is a potential strategy for mitigating the losses in NSTX-Upgrade in addition to raising the magnetic field or influencing the edge density.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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↗
Classical Impurity Ion Confinement in a Toroidal Magnetized Fusion Plasma
Explore the source record for details and available documents.
Kumar, STA↗
Introduction to the special topic on inertial confinement fusion diagnostics
In the first decade of the 21st century, the United States began operating three leading Inertial Confinement Fusion (ICF) and High-Energy Density Physics (HEDP) facilities. The National Ignition Facility (NIF) began its quest to first surpass the Lawson Criterion and then reach thermonuclear ignition and thermonuclear gain. The Omega EP (Extended Performance) laser provides unprecedented energy, reliability, and consistency for petawatt (PW)-class laser experiments. The Omega laser continues to support over 1000 ICF and HEDP experiments each year. The Z pulsed power machine is a Z-pinch plasma confinement device that was refurbished to increase current drive and improve reliability. The extreme states of matter created in these new facilities required advancement of many technologies. The most important among these was the significant improvement in instrumentation capabilities needed to diagnose plasma conditions and resulting nuclear products. The instruments needed a better temporal, spatial, and energy resolution. They also needed to be hardened against EMP and radiation. Typical events in an ICF implosion include compressing the deuterium–tritium (DT) fuel to greater than 60 g/cm 3 and heating the DT to a temperature greater than 10 keV (110 000 000 K). The DT atoms react to produce 14 MeV neutrons from a volume with a radius of ~30 μm. When the fuel ignites, the DT burn prop agates into cold fuel and the neutron-emitting region can reach sizes of over 200 μm. Furthermore, the burn typically lasts less than 150 ps, but the burn duration will decrease as the DT yield increases. Burn durations as short as 90 ps have been recorded and are predicted to decrease still further to 20 ps. X-ray imaging diagnostics are needed to measure the acceleration and velocity of the capsule enclosing the DT, where the implosion velocities can reach over 400 km/s.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
ThinCurr: An open-source 3D thin-wall eddy current modeling code for the analysis of large-scale systems of conducting structures
In this paper we present a new thin-wall eddy current modeling code, ThinCurr, for studying inductively-coupled currents in 3D conducting structures -- with primary application focused on the interaction between currents flowing in coils, plasma, and conducting structures of magnetically-confined plasma devices. The code utilizes a boundary finite element method on an unstructured, triangular grid to accurately capture device structures. The new code, part of the broader Open FUSION Toolkit, is open-source and designed for ease of use without sacrificing capability and speed through a combination of Python, Fortran, and C/C++ components. Scalability to large models is enabled through use of hierarchical off-diagonal low-rank compression of the inductance matrix, which is otherwise dense. Ease of handling large models of complicated geometry is further supported by automatic determination of supplemental elements through a greedy homology approach. Here, a detailed description of the numerical methods of the code and verification of the implementation of those methods using cross-code comparisons against the VALEN code and Ansys commercial analysis software is shown.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Generalized charge-screening in relativistic Thomas–Fermi model
In this paper, we study the charge shielding within the relativistic Thomas-Fermi model for a wide range of electron number-densities and the atomic-number of screened ions. A generalized energy-density relation is obtained using the force-balance equation and taking into account the Chandrasekhar's relativistic electron degeneracy pressure. By numerically solving a second-order nonlinear differential equation, the Thomas-Fermi screening length is investigated, and the results are compared for three distinct regimes of the solid-density, warm-dense-matter, and white-dwarfs (WDs). It is revealed that our nonlinear screening theory is compatible with the exponentially decaying Thomas-Fermi-type shielding predicted by the linear response theory. Moreover, the variation of relative Thomas-Fermi screening length shows that extremely dense quantum electron fluids are relatively poor charge shielders. Calculation of the total number of screening electrons around a nucleus shows that there is a position of maximum number of screening localized electrons around the screened nucleus, which moves closer to the point-like nucleus by increase in the plasma number density but is unaffected due to increase in the atomic-number value. It is discovered that the total number of screening electrons, (N{sub s}∝r{sub TF}{sup 3}/r{sub d}{sup 3} where r{sub TF} and r{sub d} are the Thomas-Fermi and interparticle distance, respectively) has a distinct limit for extremely dense plasmas such as WD-cores and neutron star crusts, which is unique for all given values of the atomic-number. This is equal to saying that in an ultrarelativistic degeneracy limit of electron-ion plasma, the screening length couples with the system dimensionality and the plasma becomes spherically self-similar. Current analysis can provide useful information on the effects of relativistic correction to the charge screening for a wide range of plasma density, such as the inertial-confined plasmas and compact stellar objects.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
A scaling law of the neutral penetration length and Balmer- α wing shape in high-temperature plasmas
Hydrogen atoms penetrating deep inside high-temperature magnetically confined plasmas by repetitive charge-exchange collisions result in a particle source of the plasma, which affects the plasma transport significantly. In this paper, we present an approximate solution of the fluid equations for neutral transport and an analytical representation of the neutral penetration length, in a simplified plasma geometry. This analysis predicts a power-law decay in the Balmer-α line wings which reflects the velocity distribution of the neutral atoms, with the power-law exponent analytically represented as well. These scaling laws are compared with a simple Monte–Carlo simulation and spectroscopic observations of Large Helical Device plasmas. Since the Balmer-α line wings are experimentally accessible, our formulation opens the possibility to quickly estimate the neutral penetration length from spectroscopic observations.
neutral opacity↗
FY2018 Fusion Theory and Simulation Milestone on Plasma Materials Interaction (PMI) (Final Report)
The interaction of the boundary plasma with the material surfaces in magnetically confined plasmas is among the most critical problems in fusion energy science. In FY2018, perform high performance computing simulations with coupled boundary plasma physics and materials surface models to predict the fuel recycling and tritium retention of the ITER divertor for D-T burning plasma conditions, accounting for erosion, re-deposition and impurity transport in the plasma boundary, and an initial evaluation of the influence of material deposition on the recycling and retention.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Experimental study of spectral index, mode coupling, and energy cascading in a turbulent, hot-ion plasma.
Experiments are described in which the spectrum of electrostatic potential fluctuations was measured with a capacitive probe in a steady state turbulent plasma confined in a magnetic mirror geometry. The plasma electron concentration ranged between 5 x 10 to the seventh power and 5 x 10 to the eighth power, the electron temperature ranged between 8 and 38 eV, the ion temperature between 350 and 960 eV, and the maximum-magnetic field was 1.0 T. When oscillation peaks were absent, the amplitude spectrum of electrostatic potential fluctuations was found to have a power law dependence on the frequency. The value of the spectral index was influenced by the probe position in the field, and at a certain position, by the plasma characteristics. The spectral index was generally below the value of 2.5 predicted by turbulence theories formulated in k space. The effects on the amplitude spectrum of a sinusoidal external modulation of the anode voltage were assessed over the range between 0 and 100 kHz. The spectra exhibited several mode coupling phenomena, including enhanced harmonics of the externally imposed frequency, and sideband modulation of plasma generated oscillation peaks.
Roth, J. R.↗
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↗
Internal transport barrier triggered by non-linear lower hybrid wave deposition under condition of beam-driven toroidal rotation
By using gyro-Landau fluid transport model (GLF23), time-dependent integrated modeling is carried out using TRANSP to explore the dynamic process of internal transport barrier (ITB) formation in the neutral beam heating discharges. When the current profile is controlled by LHCD (lower hybrid current drive), with appropriate neutral beam injection, the nonlinear interplay between the transport determined gradients in the plasma temperature (T{sub i,e}) and toroidal velocity (V{sub ϕ}) and the E×B flow shear (including q-profile) produces transport bifurcations, generating spontaneously a stepwise growing ITB. In the discharge, the constraints imposed by the wave propagation condition causes interplay of the LH driven current distribution with the plasma configuration modification, which constitutes non-linearity in the LH wave deposition. The non-linear effects cause bifurcation in LHCD, generating two distinct quasi-stationary reversed magnetic shear configurations. The change of current profile during the transition period between the two quasi-stationary states results in increase of the E×B shearing flow arising from toroidal rotation. The turbulence transport suppression by sheared E×B flow during the ITB development is analysed, and the temporal evolution of some parameters characterized the plasma confinement is examined. Ample evidence shows that onset of the ITB development is correlated with the enhancement of E×B shearing rate caused by the bifurcation in LHCD. It is suggested that the ITB triggering is associated with the non-linear effects of the LH power deposition.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
The complete set of Casimirs in Hall-magnetohydrodynamics
A procedure for determining all the Casimir constants of motion in magnetohydrodynamics (MHD) [E. Hameiri, Phys. Plasmas 11, 3423 (2004)] is extended to Hall-MHD. We obtain and solve differential equations for the variational derivatives of all the Casimirs, which must be satisfied for any dynamically accessible motion in Hall-MHD. In an extension of the more commonly considered Hall-MHD model, we also include the electron fluid entropy. The most interesting case for plasma confinement, which is usually true for axisymmetric configurations but desirable in general, is when both the magnetic field and the ion velocity field form the two separate families of nested toroidal surfaces. The Casimirs are then three functionals for each surface, involving the fluxes of certain vector fields and the number of particles contained in each. We also determine a family of independent Casimirs in a general configuration.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
A model for particle confinement in a toroidal plasma subject to strong radial electric fields
A toroidal plasma is confined and heated by the simultaneous application of strong d.c. magnetic fields and electric fields. Strong radial electric fields (about 1 kilovolt per centimeter) are imposed by biasing the plasma with up to 12 negative electrode rings which surround its minor circumference. The plasma containment is consistent with a balance of two processes: a radial infusion of ions in those sectors not containing electrode rings, resulting from the radially inward electric fields; and ion losses to the electrode rings, each of which acts as a sink and draws ions out the plasma in the manner of a Langmuir probe in the ion saturation regime. The highest density on axis which has been observed so far in this steady-state plasma is 6.2 x 10 to the 12th power particles per cubic centimeter, for which the particle containment time is 2.5 milliseconds. The deuterium ion kinetic temperature for these conditions was in the range of 360 to 520 eV.
Roth, J. R.↗
Resonance Overlap Axial Trapping and Magnetic Field Scaling in Asymmetry Induced Transport (Final Technical Report, Revised)
The research reported here involves studies of radial particle transport in a cylindrical, low-density Malmberg-Penning non-neutral plasma trap. The research is primarily experimental but involves careful comparisons to analytical theory and includes the results of a single-particle computer code. The transport is produced by applied electric fields that break the cylindrical symmetry of the trap, hence the term ``asymmetry-induced transport.'' Our computer studies have revealed the importance of a previously ignored class of particles that become trapped in the asymmetry potential. In many common situations these particles exhibit large radial excursions and dominate the radial transport. On the experimental side, we have developed new data analysis techniques that allowed us to determine the magnetic field dependence of the transport and to place empirical constraints on the form on the transport equation. Experiments designed to test the computer code results gave varying degrees of agreement with further work being necessary to understand the results. This work expands our knowledge of the varied mechanisms of cross-magnetic-field transport and should be of use to other workers studying plasma confinement.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Plans for Liquid Metal Divertor in Tokamak Compass
The COMPASS tokamak (R = 0.56 m, a = 0.2 m, B{sub T} = 1.3 T, I{sub p} ~ 300 kA, pulse duration 0.4 s) operates in ITER-like plasma shape in H-mode with Type-I ELMs. In 2019, we plan to install into the divertor a test target based on capillary porous system filled with liquid lithium/tin. This single target will be inclined toroidally in order to be exposed to ITER-relevant surface heat flux (20 MW/m{sup 2}). Based on precisely measured actual heat fluxes, our simulations predict (for 45° inclination, without accounting for the lithium vapor shielding) the surface temperature rises up to 700°C within 120 ms of the standard ELMy H-mode heat flux with ELM filaments reaching hundreds MW/m{sup 2}. Significant lithium vaporization is expected. The target surface will be observed by spectroscopy, fast visible and infrared cameras. The scientific program will be focused on operational issues (redeposition of the evaporated metal, ejection of droplets, if any) as well as on the effect on the plasma physics (improvement of plasma confinement, L–H power threshold, Z{sub eff}, etc.). After 2024, a closed liquid divertor may be installed into the planned COMPASS Upgrade tokamak (R = 0.84 m, a = 0.3 m, B{sub T} = 5 T, I{sub p} = 2 MA, P{sub in} = 8 MW, pulse duration ~2 s) with ITER-relevant heat fluxes loading the entire toroidal divertor.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Analytical solutions for Tokamak equilibria with reversed toroidal current
In tokamaks, an advanced plasma confinement regime has been investigated with a central hollow electric current with negative density which gives rise to non-nested magnetic surfaces. We present analytical solutions for the magnetohydrodynamic equilibria of this regime in terms of non-orthogonal toroidal polar coordinates. These solutions are obtained for large aspect ratio tokamaks and they are valid for any kind of reversed hollow current density profiles. The zero order solution of the poloidal magnetic flux function describes nested toroidal magnetic surfaces with a magnetic axis displaced due to the toroidal geometry. The first order correction introduces a poloidal field asymmetry and, consequently, magnetic islands arise around the zero order surface with null poloidal magnetic flux gradient. An analytic expression for the magnetic island width is deduced in terms of the equilibrium parameters. We give examples of the equilibrium plasma profiles and islands obtained for a class of current density profile.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗