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All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Scientific and Computational Challenges of the Fusion Simulation Program (FSP)

This paper highlights the scientific and computational challenges facing the Fusion Simulation Program (FSP) a major national initiative in the United States with the primary objective being to enable scientific discovery of important new plasma phenomena with associated understanding that emerges only upon integration. This requires developing a predictive integrated simulation capability for magnetically-confined fusion plasmas that are properly validated against experiments in regimes relevant for producing practical fusion energy. It is expected to provide a suite of advanced modeling tools for reliably predicting fusion device behavior with comprehensive and targeted science-based simulations of nonlinearly-coupled phenomena in the core plasma, edge plasma, and wall region on time and space scales required for fusion energy production. As such, it will strive to embody the most current theoretical and experimental understanding of magnetic fusion plasmas and to provide a living framework for the simulation of such plasmas as the associated physics understanding continues to advance over the next several decades. Substantive progress on answering the outstanding scientific questions in the field will drive the FSP toward its ultimate goal of developing the ability to predict the behavior of plasma discharges in toroidal magnetic fusion devices with high physics fidelity on all relevant time and space scales. From a computational perspective, this will demand computing resources in the petascale range and beyond together with the associated multi-core algorithmic formulation needed to address burning plasma issues relevant to ITER - a multibillion dollar collaborative experiment involving seven international partners representing over half the world's population. Even more powerful exascale platforms will be needed to meet the future challenges of designing a demonstration fusion reactor (DEMO). Analogous to other major applied physics modeling projects (e.g., Climate Modeling), the FSP will need to develop software in close collaboration with computers scientists and applied mathematicians and validated against experimental data from tokamaks around the world. Specific examples of expected advances needed to enable such a comprehensive integrated modeling capability and possible "co-design" approaches will be discussed. __________________________________________________

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Axial confinement of wire array Z-pinch precursor plasmas by a pulsed magnetic mirror field

A novel wire-array Z-pinch platform has been developed to study the effect of a pulsed magnetic mirror field on a collisional high energy density plasma. The mirror is driven in series with the Z-pinch target and, therefore, scales with driver current. Axial confinement is observed based on visible imaging and axial flow velocity measurements. The presence of axial compressing J xB force is determined indirectly based on Thomson scattering and interferometry measurements and corroborated by three-dimensional extended-MHD simulations. Compared to non-magnetized wire array Z-pinch, a modified pulsed mirror configuration is observed to increase precursor plasma thermal energy density by about 30%. If optimized, such a configuration could potentially improve magnetized liner inertial fusion performance by reducing axial plasma end loss.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

2014 US TTF Meeting Summary

The ultimate goal of the U.S. Transport Task Force is to develop a physics-based understanding of confinement and particle, momentum and heat transport in magnetic fusion devices. This understanding should be of sufficient depth that it allows the development of predictive models of plasma transport that can be validated against experiment, and then used to anticipate the future performance of burning plasmas in ITER, as well as to provide guidance to the design of next-step fusion nuclear science facilities. To achieve success in transport science, it is essential to characterize local fluctuations and transport in toroidal plasmas, to understand the basic mechanisms responsible for transport, and ultimately to control these transport processes. These goals must be pursued in multiple areas, including ion and electron thermal transport, particle and momentum transport, the physics of H-modes and the edge pedestal, Internal Transport Barriers, energetic particle transport and 3D effects on all the underlying transport processes. Demonstrating our understanding requires multiple, successful, quantitative tests of theory, simulation and modeling using experimental results in fusion-relevant and basic plasmas (i.e., verification and validation). The 2014 U.S. TTF meeting was held in April 2014 in San Antonio TX to provide a forum for leading scientists focused on the study of transport of particles, momentum and heat in fusion plasmas. Approximately 110 scientists from the US and several from the EU and from China attended and heard oral talks on recent transport results. Several poster sessions were also held. One day of plenary talks were followed by Breakout sessions and poster sessions that were held on focused topics, including L-H transition physics, energetic particles, transport in high performance plasmas, divertor particle and heat flux management and innovative divertor designs, fundamental turbulence studies, end edge transport shortfall.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Three-dimensional modeling of the neutron spectrum to infer plasma conditions in cryogenic inertial confinement fusion implosions

Neutron-based diagnostics are typically used to infer compressed core conditions such as areal density and ion temperature in deuterium–tritium (D–T) inertial confinement fusion (ICF) implosions. Asymmetries in the observed neutron-related quantities are important to understanding failure modes in these implosions. Neutrons from fusion reactions and their subsequent interactions including elastic scattering and neutron-induced deuteron breakup reactions are tracked to create spectra. Here, it is shown that background subtraction is important for inferring areal density from backscattered neutrons and is less important for the forward-scattered neutrons. A three-dimensional hydrodynamic simulation of a cryogenic implosion on the OMEGA Laser System [T. R. Boehly et al., Opt. Commun. 133, 495 (1997)] using the hydrodynamic code HYDRA [M. M. Marinak et al., Phys. Plasmas 8, 2275 (2001)] is post-processed using the tracking code IRIS3D. It is shown that different parts of the neutron spectrum from the view can be mapped into different regions of the implosion, enabling an inference of an areal-density map. It is also shown that the average areal-density and an areal-density map of the compressed target can be reconstructed with a finite number of detectors placed around the target chamber. Ion temperatures are inferred from the width of the D–D and D–T fusion neutron spectra. Backgrounds can significantly alter the inferred ion temperatures from the D–D reaction, whereas they insignificantly influence the inferred D–T ion temperatures for the areal densities typical of OMEGA implosions. Asymmetries resulting in fluid flow in the core are shown to influence the absolute inferred ion temperatures from both reactions, although relative inferred values continue to reflect the underlying asymmetry pattern. The work presented here is part of the wide range of the first set of studies performed with IRIS3D. Finally, this code will continue to be used for post-processing detailed hydrodynamic simulations and interpreting observed neutron spectra in ICF implosions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

The Princeton Field-Reversed Configuration for Compact Nuclear Fusion Power Plants

The Princeton Field-Reversed Configuration (PFRC) nuclear fusion reactor concept is an innovative approach to fusion power generation prioritizing low neutron production and small size. Here, a combination of analytical modeling and numerical simulation shows that the novel heating approach generates an FRC with closed field lines. Simulation data from a single-particle Hamiltonian code predicts ms-scale plasma heating in reactor-scale conditions while PIC codes predict formation of warm FRC plasmas from initial mirror fields. The PFRC-1 and PFRC-2 experiments have heated electrons to energies well in excess of 100 eV and plasma durations to 300 ms, more than 10 4 times longer than the predicted tilt instability growth time. From these data, we have created a development plan and anticipated performance metrics for a fusion reactor based on the PFRC concept. The resulting 1–10 MW PFRC reactors would be suitable for diverse applications, from submarines to urban environments to space propulsion. PFRC is a steady-state, driven magnetic confinement device. Plasma, inside a cylindrical array of coils, is confined and heated by external RF antennae. PFRC would be ultra-low radiation due to both its fuel and small size. The choice of advanced fuels, deuterium and helium-3 (D– 3 He), may be enabled by the high-β FRC configuration. The small size of the reactor would enable rapid exhaust of the dangerous tritium ash. Low radiation would make the reactor safer to operate and, in combination with simple geometry and small size, dramatically lowers development and maintenance costs. This review paper gives an introduction to the physics of the PFRC and a summary of the PFRC-2 experiment results to date. It then discusses the future program plan and how PFRC reactors would be commercialized.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Liquid lithium as divertor material to mitigate severe damage of nearby components during plasma transients

The successful operation of thermonuclear fusion reactors such as ITER, DEMO, and future commercial plants is mainly determined by the optimum choice of materials for various components. The objective of this work is to accurately and comprehensively simulate the entire device in 3D to predict pros and cons of various materials, e.g., liquid lithium in comparison to tungsten and carbon to predict future ITER-like and DEMO divertor performances. We used our comprehensive HEIGHTS simulation package to investigate ITER-like components response during transient events in exact 3D geometry. Starting from the lost hot core plasma particles through SOL, deposition on the divertor surface, and the generation of secondary plasma of divertor materials. Our simulations predicted significant reduction in the heat loading and damage to the divertor nearby and internal components in the case when lithium is used on the divertor plates. While if tungsten or carbon are used on the divertor plate, significant melting areas and vaporization spots can occur (less for carbon) on the reflector, dome, and stainless steel tubes, and even parts of the first walls can melt due to the high radiation power of the secondary divertor plasma. Lithium photon radiation deposition into the divertor and nearby surfaces was decreased by two orders of magnitude compared to tungsten and by one order of magnitude compared to carbon. This analysis showed that using liquid lithium for ITER-like surfaces and future DEMO can lead to significant enhancement in components lifetime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Experimental study of parametric dependence of electron-scale turbulence in a spherical tokamak

Electron-scale turbulence is predicted to drive anomalous electron thermal transport. However, experimental study of its relation with transport is still in its early stage. On the National Spherical Tokamak Experiment (NSTX), electron-scale density fluctuations are studied with a novel tangential microwave scattering system with high radial resolution of {+-}2 cm. Here, we report a study of parametric dependence of electron-scale turbulence in NSTX H-mode plasmas. The dependence on density gradient is studied through the observation of a large density gradient variation in the core induced by an edge localized mode (ELM) event, where we found the first clear experimental evidence of density gradient stabilization of electron-gyro scale turbulence in a fusion plasma. This observation, coupled with linear gyro-kinetic calculations, leads to the identification of the observed instability as toroidal electron temperature gradient (ETG) modes. It is observed that longer wavelength ETG modes, k{sub Up-Tack }{rho}{sub s} Less-Than-Or-Equivalent-To 10 ({rho}{sub s} is the ion gyroradius at electron temperature and k{sub Up-Tack} is the wavenumber perpendicular to local equilibrium magnetic field), are most stabilized by density gradient, and the stabilization is accompanied by about a factor of two decrease in electron thermal diffusivity. Comparisons with nonlinear ETG gyrokinetic simulations show ETG turbulence may be able to explain the experimental electron heat flux observed before the ELM event. The collisionality dependence of electron-scale turbulence is also studied by systematically varying plasma current and toroidal field, so that electron gyroradius ({rho}{sub e}), electron beta ({beta}{sub e}), and safety factor (q{sub 95}) are kept approximately constant. More than a factor of two change in electron collisionality, {nu}{sub e}{sup *}, was achieved, and we found that the spectral power of electron-scale turbulence appears to increase as {nu}{sub e}{sup *} is decreased in this collisonality scan. However, both linear and nonlinear simulations show no or weak dependence with the electron-ion collision frequency, {nu}{sup e/i}. Instead, other equilibrium parameters (safety factor, electron density gradient, for example) affect ETG linear growth rate and electron thermal transport more than {nu}{sup e/i} does. Furthermore, electron heat flux predicted by the simulations is found to have an order-of-magnitude spatial variation in the experimental measurement region and is also found to be much smaller than experimental levels except at one radial location we evaluated. The predicted electron heat flux is shown to be strongly anti-correlated with density gradient, which varies for a factor of three in the measurement region, which is in agreement with the density gradient dependence study reported in this paper.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

H-mode characterisation for dominant ECRH and comparison to dominant NBI or ICRF heating at ASDEX Upgrade

At ASDEX Upgrade the ECRH system has been upgraded to provide up to 4 MW of heating power at 140 GHz (or 2.2 MW at 105 GHz). The power at 140 GHz exceeds the minimum H-mode power threshold for typical high I{sub p}, B{sub t} conditions by approximately a factor of two. The upgrade allows H-modes with dominant electron heating and significant electron-ion heat exchange to be studied, i.e. the situation expected in ITER. This paper reports on systematic studies varying the heating mix with NBI, ICRF and ECRH and its effect on pedestal parameters and core transport. The H-mode pedestal is hardly affected by the choice of heating mix, but the ion temperature in the plasma center is found to vary significantly. The ion channel dominates heat transport and ion temperature gradient modes (ITG) are found to be the most unstable microinstability in all the scenarios considered. R/L{sub Ti} at half radius reduces by a factor of two when T{sub e}/T{sub i} increases from 0.9 to 1.5. TGLF modelling of the electron and ion temperature and electron density profiles shows very good agreement with the experimental data when applying a realistic sawtooth model.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

Tungsten impurity transport experiments in Alcator C-Mod to address high priority research and development for ITER

Experiments in Alcator C-Mod tokamak plasmas in the Enhanced D-alpha H-mode regime with ITER-like mid-radius plasma density peaking and Ion Cyclotron Resonant heating, in which tungsten is introduced by the laser blow-off technique, have shown that accumulation of tungsten in the central region of the plasma does not take place in these conditions. The measurements obtained are consistent with anomalous transport dominating tungsten transport except in the central region of the plasma where tungsten transport is neoclassical, as previously observed in other devices with dominant neutral beam injection heating, such as JET and ASDEX Upgrade. In contrast to such results, yet, the measured scale lengths for plasma temperature and density in the central region of these Alcator C-Mod plasmas, with density profiles relatively flat in the core region due to the lack of core fuelling, are favourable to prevent inter and intra sawtooth tungsten accumulation in this region under dominance of neoclassical transport. Simulations of ITER H-mode plasmas, including both anomalous (modelled by the Gyro-Landau-Fluid code GLF23) and neoclassical transport for main ions and tungsten and with density profiles of similar peaking to those obtained in Alcator C-Mod show that accumulation of tungsten in the central plasma region is also unlikely to occur in stationary ITER H-mode plasmas due to the low fuelling source by the neutral beam injection (injection energy ~ 1 MeV), which is in consistent agreement with findings in the Alcator C-Mod experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Development of long pulse RF heating and current drive for H-mode scenarios with metallic walls in WEST

The longstanding expertise of the Tore Supra team in long pulse heating and current drive with radiofrequency (RF) systems will now be exploited in the WEST device (tungsten-W Environment in Steady-state Tokamak) [1]. WEST will allow an integrated long pulse tokamak programme for testing W-divertor components at ITER-relevant heat flux (10-20 MW/m{sup 2}), while treating crucial aspects for ITER-operation, such as avoidance of W-accumulation in long discharges, monitoring and control of heat fluxes on the metallic plasma facing components (PFCs) and coupling of RF waves in H-mode plasmas. Scenario modelling using the METIS-code shows that ITER-relevant heat fluxes are compatible with the sustainment of long pulse H-mode discharges, at high power (up to 15 MW / 30 s at I{sub P} = 0.8 MA) or high fluence (up to 10 MW / 1000 s at I{sub P} = 0.6 MA) [2], all based on RF heating and current drive using Ion Cyclotron Resonance Heating (ICRH) and Lower Hybrid Current Drive (LHCD). This paper gives a description of the ICRH and LHCD systems in WEST, together with the modelling of the power deposition of the RF waves in the WEST-scenarios.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2018

Impact of magnetic islands in the plasma edge on particle fueling and exhaust in the HSX and W7-X stellarators

The edge magnetic structure in the Helically Symmetric eXperiment (HSX) and Wendelstein 7X (W7-X) stellarators has been shown to have a significant impact on the particle fueling and exhaust of the plasma main species (hydrogen) as well as impurity helium. For HSX, the plasma sourcing to exhaust ratio, quantified by the effective and global particle confinement times τ p * and τ p , H , respectively, increases when a magnetic island chain is located in the plasma edge. The fueling efficiency is reduced by 25% when the plasma boundary is deformed by the magnetic islands. The X-point geometry also yields higher plasma temperatures in front of the main recycling region. When the island is moved radially inward, both τ p * and τ p decrease by 10 % – 25 % depending on plasma density. The τ p , H results rely heavily on EMC3-EIRENE modeling which confirms reduced fueling efficiency due to more rapid ionization in the outward shifted island position. These findings suggest that for a helically optimized system like HSX, the plasma fueling from the recycling source, as well as from active gas injection, can be controlled by the magnetic island chain in the plasma edge—which is a basic requirement for a divertor system. This process is also effective for the control of effective helium exhaust times, as τ p , H e * measured by perturbative gas puff experiments is reduced by up to 40% when the islands are shifted inwards. For Wendelstein 7-X, a similar reduction of τ p , H e * was inferred when magnetic islands were moved from the far plasma edge into the confined plasma region. Finally, however, the effective confinement features of H as the main plasma species were not affected due to the non-optimal position of the magnetic islands with respect to the highly localized ionization domain during the limiter startup campaign.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Dynamics of laser-blow-off induced Li plume in confined geometry

Dynamics of Li plasma plume created by laser-blow-off technique in air ambient is reported. Plasma plume dynamics and its optical emission are investigated in planar and confined geometries using time resolved shadowgraph imaging and optical emission spectroscopy. Significant differences in the plasma characteristics in confined geometry are quantitatively investigated by comparing the plasma parameters (temperature and density) in free expansion and confined geometry configurations. Dynamics and physical parameters of the primary as well as the reflected shock waves (in confined geometry) and their interactions with expanding plasma are briefly addressed. A large enhancement in the emission intensities of Li I 610.3 nm (2p {sup 2}P{sub 1/2,3/2}← 3d {sup 2}P{sub 3/2,5/2}) and 670.8 nm (2s {sup 2}S{sub 1/2}← 2p {sup 2}P{sub 1/2,3/2}) is correlated with the shock wave dynamics in the two geometries. Strong self reversal in the neutral emission infers an increase in the population density of neutrals within the confined plasma plume.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Merging of the superbanana plateau and $\sqrt{v}$ transport regimes in nearly quasisymmetric stellarators

We report alpha particle confinement is one of the most demanding issues for stellarators. It now seems clear that it is possible to design optimized stellarators that confine the background plasma at near tokamak radial transport levels. Moreover, adequate collisionless alpha particle confinement is possible in the core of a highly optimized stellarator. Here, the collisional confinement of barely trapped alphas in an optimized stellarator is considered by accounting for the resonance due to the reversal in direction of the drift within a flux surface and investigating the sensitive role of magnetic shear in keeping this resonance close to the passing boundary in some nearly quasisymmetric stellarator configurations. The treatment relies on a narrow collisional boundary layer formulation that combines the responses of both these resonant pitch angle alphas and the remaining barely trapped alphas. A novel merged regime treatment leads to explicit expressions for the energy diffusivity for both superbanana plateau (or resonant plateau) and $\sqrt{v}$ transport in the large aspect ratio limit for a slowing down tail alpha distribution function, where v is the effective pitch angle scattering collision frequency of the trapped alphas off the background ions. Depending on the details of the optimization scheme and the sign of the magnetic shear, modest magnetic shear can be used to reduce superbanana (or resonant) plateau transport to below the $\sqrt{v}$ transport level. In addition, a quasilinear equation retaining spatial diffusion is derived for a general alpha distribution function that allows the radial alpha transport to modify the distribution so it is no longer isotropic in velocity space.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Diagnosing magnetized liner inertial fusion experiments on Z

Magnetized Liner Inertial Fusion experiments performed at Sandia's Z facility have demonstrated significant thermonuclear fusion neutron yields (∼10{sup 12} DD neutrons) from multi-keV deuterium plasmas inertially confined by slow (∼10 cm/μs), stable, cylindrical implosions. Effective magnetic confinement of charged fusion reactants and products is signaled by high secondary DT neutron yields above 10{sup 10}. Analysis of extensive power, imaging, and spectroscopic x-ray measurements provides a detailed picture of ∼3 keV temperatures, 0.3 g/cm{sup 3} densities, gradients, and mix in the fuel and liner over the 1–2 ns stagnation duration.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

A compact and continuously driven supersonic plasma and neutral source

A compact and repetitively driven plasma source has been developed by utilizing a magnetized coaxial plasma gun (MCPG) for diagnostics requiring deep penetration of a large amount of neutral flux. The system consists of a MCPG 95mm in length with a DN16 ConFlat connection port and an insulated gate bipolar transistor (IGBT) inverter power unit. The power supply consists of an array of eight IGBT units and is able to switch the discharge on and off at up to 10 kV and 600 A with a maximum repetitive frequency of 10 kHz. Multiple short duration discharge pulses maximize acceleration efficiency of the plasmoid. In the case of a 10 kHz operating frequency, helium-plasmoids in the velocity range of 20 km/s can be achieved.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Solid debris collection for radiochemical diagnostics at the National Ignition Facility

Radiochemical analysis of post-ignition debris inside the National Ignition Facility (NIF) target chamber can help determine various diagnostic parameters associated with the implosion efficiency of the fusion capsule. This technique is limited by the ability to distinguish ablator material from other debris and by the collection efficiency of the capsule debris after implosion. Prior to designing an on-line collection system, the chemical nature and distribution of the debris inside the chamber must be determined. The focus of our current work has been on evaluating capture of activated Au hohlraum debris on passive foils (5 cm diameter, 50 cm from target center) post-shot. Preliminary data suggest that debris distribution is locally heterogeneous along the equatorial and polar line-of-sights.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2014

Lower hybrid current drive and ion cyclotron range of frequencies heating experiments in H-mode plasmas in Experimental Advanced Superconducting Tokomak

An ion cyclotron range of frequencies (ICRF) system with power up to 6.0 MW and a lower hybrid current drive (LHCD) system up to 4 MW have been applied for heating and current drive experiments in Experimental Advanced Superconducting Tokomak (EAST). Significant progress has been made with ICRF heating and LHCD for realizing the H-mode plasma operation in EAST. During 2010 and 2012 experimental campaigns, ICRF heating experiments were carried out at the fixed frequency of 27MHz, achieving effective ions and electrons heating with the H minority heating (H-MH) mode. The H-MH mode produced good plasma performance, and realized H-mode using ICRF power alone in 2012. In 2010, H-modes were generated and sustained by LHCD alone, where lithium coating and gas puffing near the mouth of the LH launcher were applied to improve the LHCD power coupling and penetration into the core plasmas of H-modes. In 2012, the combination of LHCD and ICRH power extended the H-mode duration up to over 30 s. H-modes with various types of edge localized modes (ELMs) have been achieved with H{sub IPB98}(y, 2) ranging from 0.7 to over unity. A brief overview of LHCD and ICRF Heating experiment and their application in achieving H-mode operation during these two campaigns will be presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗