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At least 1,207 records · Page 67

Cross-beam energy transfer saturation by ion trapping-induced detuning

The performance of direct-drive inertial confinement fusion implosions relies critically on the coupling of laser energy to the target plasma. Cross beam energy transfer (CBET), the resonant exchange of energy between intersecting laser beams mediated by ponderomotively driven ion-acoustic waves (IAW), inhibits this coupling by scattering light into unwanted directions. Furthermore, the variety of beam intersection angles and varying plasma conditions in an implosion results in IAWs with a range of phase velocities. Here we show that CBET saturates through a resonance detuning that depends on the IAW phase velocity and that results from trapping-induced modifications to the ion distribution functions. For smaller phase velocities, the modifications to the distribution functions can rapidly thermalize in the presence of mid-Z ions, leading to a blueshift in the resonant frequency. For larger phase velocities, the modifications can persist, leading to a redshift in the resonant frequency. Ultimately, these results may reveal pathways towards CBET mitigation and inform reduced models for radiation hydrodynamics codes to improve their predictive capability.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Application and Continued Development of Thin Faraday Collectors as a Lost Ion Diagnostic for Tokamak Fusion Plasmas

This report summarizes the accomplishment of sixteen years of work toward the development of thin foil Faraday collectors as a lost energetic ion diagnostic for high temperature magnetic confinement fusion plasmas. Following initial, proof of principle accelerator based studies, devices have been tested on TFTR, NSTX, ALCATOR, DIII-D, and JET (KA-1 and KA-2). The reference numbers refer to the attached list of publications. The JET diagnostic KA-2 continues in operation and hopefully will provide valuable diagnostic information during a possible d-t campaign on JET in the coming years. A thin Faraday foil spectrometer, by virtue of its radiation hardness, may likewise provide a solution to the very challenging problem of lost alpha particle measurements on ITER and other future burning plasma machines.

43 PARTICLE ACCELERATORS↗

Mitigation of upward and downward vertical displacement event heat loads with upper or lower massive gas injection in DIII-D

Intentionally triggered upward and downward vertical displacement events (VDEs) leading to disruptions were pre-emptively mitigated with neon massive gas injection (MGI) coming from either above or below the plasma. Global indicators of disruption mitigation effectiveness (conducted heat loads, radiated power, and vessel motion) do not show a clear improvement when mitigating with the gas jet located closer to the VDE impact area. A clear trend of improved mitigation is observed for earlier MGI timing relative to the VDE impact time. The plasma edge magnetic perturbation is seen to lock to a preferential phase during the VDE thermal quench, but this phase is not clearly matched by preliminary attempts to fit to the conducted heat load phase. Clear indications of plasma infra-red (IR) emission are observed both before and during the disruptions. This IR emission can affect calculation of disruption heat loads; here, the time decay of post-disruption IR signals is used to correct for this effect.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

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↗

A large volume uniform plasma generator for the experiments of electromagnetic wave propagation in plasma

A large volume uniform plasma generator is proposed for the experiments of electromagnetic (EM) wave propagation in plasma, to reproduce a 'black out' phenomenon with long duration in an environment of the ordinary laboratory. The plasma generator achieves a controllable approximate uniform plasma in volume of 260 mm Multiplication-Sign 260 mm Multiplication-Sign 180 mm without the magnetic confinement. The plasma is produced by the glow discharge, and the special discharge structure is built to bring a steady approximate uniform plasma environment in the electromagnetic wave propagation path without any other barriers. In addition, the electron density and luminosity distributions of plasma under different discharge conditions were diagnosed and experimentally investigated. Both the electron density and the plasma uniformity are directly proportional to the input power and in roughly reverse proportion to the gas pressure in the chamber. Furthermore, the experiments of electromagnetic wave propagation in plasma are conducted in this plasma generator. Blackout phenomena at GPS signal are observed under this system and the measured attenuation curve is of reasonable agreement with the theoretical one, which suggests the effectiveness of the proposed method.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Estimates of global recycling coefficients for LTX- β discharges

We report the first observation of global recycling coefficient R near 0.5 in the Lithium Tokamak eXperiment-β (LTX-β), significantly below the minimum R previously reported in other devices. In a series of experiments with varied Li wall conditioning, estimates of the recycling coefficient have been made using a Lyman-α array and DEGAS2 modeling. A progressive reduction in Lyman-α emission with increased lithium and an increase in edge electron temperature are observed. It is also observed that with increasing Li coating thickness, the effective particle confinement time τp* is reduced and approaches TRANSP calculated energy confinement time (τE), with τp* near τE,TRANSP for the lowest recycling coefficients. Edge temperatures approaching core plasma temperatures, first reported in LTX, can now be directly connected to estimates of the recycling coefficient and qualitatively agree with previous UEDGE simulations. The particle flux to the limiting surfaces appears to be significantly reduced in comparison with fluid scrape-off layer (SOL) models, indicating that a large fraction of the SOL ions are mirror trapped. SOL collisionality drops more than an order of magnitude below the banana regime boundary, indicating the importance of kinetic effects. Full-f 1x2v gyrokinetic simulations of SOL field lines with the GKEYLL code indicate that the fraction of ions trapped along field lines increases as collisionality drops, as a result of increased lithium evaporation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Active Divertor Heat Flux Control using Impurity Powder Dropper

Divertor plasma-facing components (PFCs) in a tokamak are typically designed to withstand average steady-state heat loads of about 5–10 MW/m 2 , a limit that applies to both solid and liquid lithium (LL) PFCs. Exceeding these design values can result in surface damage to tungsten PFCs or excessive lithium (Li) evaporation in liquid lithium divertor (LLD) PFCs. Since exceeding the divertor heat load limits has serious consequences, it is therefore prudent to develop a tool to reduce the divertor heat load and bring the heat load to within the design limit without affecting the plasma performance. Active low Z impurity injection such as Li has been suggested as a potential solution to mitigate excess heat flux as suggested previously, given that non-coronal radiation can be quite large ~ 20–30 MJ per mole of injected Li. Li is considered desirable for reducing the edge neutral recycling helping to improve plasma energy confinement. In this paper, we model the Impurity Power Dropper (IPD) to investigate its potential of divertor heat flux control. The IPD is typically located at the top of the tokamak device and uses a vertical drift tube of a few meters. In the 2 m drift tube case, the IPD powder is accelerated to ~ 6 m/sec before reaching the plasma with the upper divertor configuration, matching the condition for the in-board side pellet injection case. By modeling the IPD geometry we determined the IPD powder deposition profile, and thus the non-coronal radiation and ionization profiles in time as well. From the enhanced radiation power loss, it is therefore possible to reduce the divertor heat load using the divertor simulation code. In conclusion, the IPD divertor heat flux control can be tested in the facilities with IPD including ST-40, DIII-D, EAST, WEST and NSTX-U.

Active lithium injection↗

H-mode pedestal scaling in DIII-D, ASDEX Upgrade, and JET

Multidevice pedestal scaling experiments in the DIII-D, ASDEX Upgrade (AUG), and JET tokamaks are presented in order to test two plasma physics pedestal width models. The first model proposes a scaling of the pedestal width {Delta}/a {proportional_to} {rho}*{sup 1/2} to {rho}* based on the radial extent of the pedestal being set by the point where the linear turbulence growth rate exceeds the ExB velocity. In the multidevice experiment where {rho}* at the pedestal top was varied by a factor of four while other dimensionless parameters where kept fixed, it has been observed that the temperature pedestal width in real space coordinates scales with machine size, and that therefore the gyroradius scaling suggested by the model is not supported by the experiments. The density pedestal width is not invariant with {rho}* which after comparison with a simple neutral fuelling model may be attributed to variations in the neutral fuelling patterns. The second model, EPED1, is based on kinetic ballooning modes setting the limit of the radial extent of the pedestal region and leads to {Delta}{sub {psi} {proportional_to}} {beta}{sub p}{sup 1/2}. All three devices show a scaling of the pedestal width in normalised poloidal flux as {Delta}{sub {psi} {proportional_to}} {beta}{sub p}{sup 1/2}, as described by the kinetic ballooning model; however, on JET and AUG, this could not be distinguished from an interpretation where the pedestal is fixed in real space. Pedestal data from all three devices have been compared with the predictive pedestal model EPED1 and the model produces pedestal height values that match the experimental data well.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

N=2 ICRH of H majority plasmas in JET-ILW

Heating single ion species plasmas with ICRF is a challenging task: Fundamental ion cyclotron heating (w = w{sub ci}) suffers from the adverse polarization of the RF electric fields near the majority cyclotron resonance while second harmonic heating (w = 2w{sub ci}) typically requires pre-heating of the plasma ions to become efficient. Recently, w = 2w{sub ci} ICRF heating was tested in JET-ILW hydrogen plasmas in the absence of neutral beam injection (L-mode). Despite the lack of pre-heating, up to 6MW of ICRF power were coupled to the plasma leading to a transition to H-mode for P{sub ICRH}>5MW in most discharges. Heating efficiencies between 0.65-0.85 were achieved as a combination of the low magnetic field adopted (enhanced finite Larmor radius effects) and the deliberate slow rise of the ICRF power, allowing time for a fast ion population to gradually build-up leading to a systematic increase of the wave absorptivity. Although fast ion tails are a common feature of harmonic ICRF heating, the N=2 majority heating features moderate tail energies (<500keV) except at very low plasma densities (n{sub e0}<3x10{sup 19}/m{sup 3}), where fast H tails in the MeV range developed and fast ion losses became significant, leading to enhanced plasma wall interaction. The main results of these experiments will be reported.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Evolution of surface structure in laser-preheated, perturbed materials

Here, we report an experimental and computational study investigating the effects of laser preheat on the hydrodynamic behavior of a material layer. In particular, we find that perturbation of the surface of the layer results in a complex interaction, in which the bulk of the layer develops density, pressure, and temperature structure and in which the surface experiences instability-like behavior, including mode coupling. A uniform one-temperature preheat model is used to reproduce the experimentally observed behavior, and we find that this model can be used to capture the evolution of the layer, while also providing evidence of complexities in the preheat behavior. Lastly, this result has important consequences for inertially confined fusion plasmas, which can be difficult to diagnose in detail, as well as for laser hydrodynamics experiments, which generally depend on assumptions about initial conditions in order to interpret their results.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Role of sheared E × B flow in self-organized, improved confinement states in magnetized plasmas

A great scientific success story of magnetic fusion research in the past several decades is the theoretical development and experimental testing of the process of turbulence decorrelation and stabilization by sheared E×B flow which shows that E×B shear effects are ubiquitous in magnetized plasmas. This concept of turbulence decorrelation and stabilization has the universality needed to explain the H-mode edge transport barriers seen in limiter and divertor tokamaks, stellarators, and mirror machines; the broader edge transport barrier seen in VHmode plasmas; and the core transport barriers formed in tokamaks. Similar effects are seen in linear devices. These examples of confinement improvement are of considerable physical interest; it is not often that a system self-organizes to reduce transport when an additional source of free energy is applied to it. The transport decrease associated with E×B velocity shear is also of great practical benefit to fusion research, since it contributed to substantially increased fusion yield in all DT magnetic fusion experiments conducted to date. The fundamental physics involved in transport reduction is the effect of E×B shear on the growth, radial extent, and phase correlation of turbulent eddies in the plasma. The same basic transport reduction process can be operational in various portions of the plasma because there are a number of ways to change the radial electric field Er. An important secondary theme in this area is the synergistic effect of E×B velocity shear and magnetic shear. Although the E×B velocity shear appears to have an effect on broader classes of microturbulence, magnetic shear can mitigate some potentially harmful effects of E×B velocity shear and facilitate turbulence stabilization. Our present understanding in this area is the result of a multi-decade, intertwined effort in theory, modeling and diagnostic development combined with continuing experimental investigations. These experiments have clearly demonstrated that increased E×B shear causes reductions in turbulence and transport. Finally, the experimental findings are generally consistent with the basic theoretical models although considerable work remains to be done before we have a fully predictive theory of transport in magnetized plasmas including E×B shear effects.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Ion cyclotron resonance frequency heating in JET during initial operations with the ITER-like wall

In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall materials on the JET Ion Cyclotron Resonance Frequency (ICRF) operation is assessed and some important properties of JET plasmas heated with ICRF are highlighted. A ∼ 20% reduction of the antenna coupling resistance is observed with the ILW as compared with the JET carbon (JET-C) wall. Heat-fluxes on the protecting limiters close the antennas, quantified using Infra-Red thermography (maximum 4.5 MW/m{sup 2} in current drive phasing), are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. The location of the tungsten ICRF specific source could not be identified but some experimental observations indicate that main-chamber W components could be an important impurity source: for example, the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions, and Be evaporation in the main chamber results in a strong reduction of the impurity level. In L-mode plasmas, the ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 15%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating performance; the power is typically deposited at the plasma centre while the radiation is mainly from the outer part of the plasma bulk. Application of ICRF heating in H-mode plasmas has started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core has been observed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Asymptotic quasisymmetric high-beta three-dimensional magnetohydrodynamic equilibria near axisymmetry

Quasisymmetry (QS), a hidden symmetry of the magnetic field strength, is known to support nested flux surfaces and provide superior particle confinement in stellarators. In this work, we study the ideal magnetohydrodynamic (MHD) equilibrium and stability of high-beta plasma in a large-aspect-ratio stellarator. In particular, we show that the lowest-order description of a near-axisymmetric equilibrium vastly simplifies the problem of three-dimensional quasisymmetric MHD equilibria, which can be reduced to a standard elliptic Grad–Shafranov equation for the flux function. We show that any large-aspect-ratio tokamak, deformed periodically in the vertical direction, is a stellarator with approximate volumetric QS. We discuss exact analytical solutions and numerical benchmarks. Finally, we discuss the ideal ballooning and interchange stability of some of our equilibrium configurations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

The Center for Momentum Transport and Flow Organization in Plasmas - Final Scientific Report

Overview of University of Colorado Efforts: The University of Colorado group has focused on two primary fronts during the grant period: development of a variety of multi-point diagnostic and/or imaging analysis techniques, and momentum-transport related experiments on a variety of devices (NSTX at PPPL, CSDX at UCSD, LAPD at UCLA, DIII-D at GA). Experimental work has taken advantage of several diagnostic instruments, including fast-framing cameras for imaging of electron density fluctuations (either directly or using injected gas puffs), ECEI for imaging of electron temperature fluctuations, and multi-tipped Langmuir and magnetic probes for corroborating measurements of Reynolds and Maxwell stresses. Mode Characterization in CSDX: We have performed a series of experiments at the CSDX linear device at UCSD, in collaboration with Center PI G. Tynan's group. The experiments included a detailed study of velocity estimation techniques, including direct comparisons between Langmuir probes and image-based velocimetry from fast-framing camera data. We used the camera data in a second set of studies to identify the spatial and spectral structure of coherent modes, which illuminates wave behavior to a level of detail previously unavailable, and enables direct comparison of dispersion curves to theoretical estimates. In another CSDX study, similar techniques were used to demonstrate a controlled transition from nonlinearly coupled discrete eigenmodes to fully developed broadband turbulence. The axial magnetic field was varied from 40-240 mT, which drove the transition. At low magnetic fields, the plasma is dominated by drift waves. As the magnetic field is increased, a strong potential gradient at the edge introduces an ExB shear-driven instability. At the transition, another mode with signatures of a rotation-induced Rayleigh–Taylor instability appears at the central plasma region. Concurrently, large axial velocities were found in the plasma core. For larger magnetic fields, all of the instabilities co-exist, leading to rich plasma dynamics and fully developed broadband turbulence. Edge-Turbulence and Flow Experiments in NSTX: A series of Gas Puff Imaging (GPI) observations on NSTX revealed a quasi-periodic oscillation in the plasma edge preceding the L-H transition in a limited set of neutral beam heated plasmas. These ~3 kHz flow oscillations exhibit both long wavelength and long correlation lengths, suggesting they are zonal-flow-like. The flow oscillations are strongly correlated with modulations of the level of edge turbulence, thus the system appears to undergo a predator--prey-type limit-cycle preceding the L-H transition. However, a clear trigger for the L-H transition was not observed. Reynolds stress profiles were obtained directly from image velocimetry for L-mode periods ELM-Precursor Studies in NSTX: A separate study based on NSTX-GPI data captured the two-dimensional evolution of edge-localized mode (ELM) precursors. Precursor events were observed preceding ELMs and ELM-induced H–L back-transitions in radio-frequency heated H-mode plasmas, and the growth of the precursor mode through the ELM filamentation was imaged in the plane perpendicular to the local B-field. Strong edge intensity modulations appeared to propagate in the electron diamagnetic direction while steadily drifting radially outwards. Intensity fluctuations were observed at frequencies around 20 kHz and wavenumbers of 0.05-0.2 cm -1 . Upon growing to a trigger point, precursor fluctuations were seen to form filamentary structures and move into the scrape-off layer (SOL) explosively with radial velocities peaking at 8 km/s. Once in the SOL, filaments reverse their propagation direction and travel in the ion diamagnetic direction. Edge intensity fluctuations were strongly correlated with magnetic signals from Mirnov coils, and toroidally distributed coils estimated toroidal mode numbers of n=5-10. Quantitatively similar precursors have been observed in ohmic H-mode plasmas as well, though significantly fewer events are seen in the ohmic cases and none were observed in the near-threshold NBI H-modes studied. Development of Velocimetry and Image Analysis Techniques: Along with the experiments listed above, the Colorado group has continued to explore various velocimetry techniques and their range of validity. We have developed a “linear optical flow” code, which calculates smooth velocity maps while accurately assessing local regions of high curl. This is critical for separating spatial scales of velocity behavior, and thus transport. This code has recently come on-line, and we are currently using it to revisit a number of older datasets. Additionally, we have worked on developing pattern-recognition techniques for imaging diagnostics, based on established digital image compression algorithms. This has the potential to open the analysis of turbulent plasma behavior beyond the well-trodden Fourier and wavelet approaches. Finally, we have extended several of these image-analysis routines to multiple other diagnostic sets, including GPI datasets in NSTX and ECEI imaging on DIII-D. In one study, correlation maps were used to measure the full 2-D mode structure of drift-wave level structures using ECEI for the first time. This enabled direct comparison to gyrokinetic simulations using the GEM code, from which is was determined that TEM modes were being measured in DIII-D.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Hollow-cathode chemical vapor deposition of thick, low-stress diamond-like carbon films

A radio-frequency (RF), hollow-cathode plasma source with confining magnetic field is described for the chemical vapor deposition of thick ( > 10 µm), amorphous diamond-like carbon ablator films for inertial confinement fusion applications. Plasma is characterized by optical emission spectroscopy, while properties of the resultant films are measured by a combination of profilometry, Rutherford backscattering spectrometry, elastic recoil detection analysis, X-ray diffraction, Raman spectroscopy, and atomic force microscopy. The dependence of the deposition rate, film density, elemental composition, self-bias and residual stress is reported as a function of RF power. Higher density films were found when using Ar plasma, than N 2 or H 2 plasma. The coatings produced are x-ray amorphous, exhibit low compressive stress ( ~ 100 MPa), high density ( < 1.7 g/cm 3 ), hydrogen content of ~ 30 at.%, and a low average roughness of 0.75 nm. Finally, applications of these films as tunable-density ablators for inertial confinement fusion experiments are discussed.

36 MATERIALS SCIENCE↗

Formation of Imploding Plasma Liners for HEDP and MIF Application

Plasma jets with high density and velocity have a number of important applications in fusion energy and elsewhere, including plasma refueling, disruption mitigation in tokamaks, magnetized target fusion, injection of momentum into centrifugally confined mirrors, plasma thrusters, and high energy density plasmas (HEDP). In Magneto-Inertial Fusion (MIF), for example, an imploding material liner is used to compress a magnetized plasma to fusion conditions and to confine the resulting burning plasma inertially to obtain the necessary energy gain. The imploding shell may be solid, liquid, gaseous, or a combination of these states. The presence of the magnetic field in the target plasma suppresses thermal transport to the plasma shell, thus lowering the imploding power needed to compress the target to fusion conditions. This allows the required imploding momentum flux to be generated electromagnetically using off-the-shelf pulsed power technology. Practical schemes for standoff delivery of the imploding momentum flux are required and are open topics for research. One approach for accomplishing this, called plasma jet driven magneto-inertial fusion (PJMIF), uses a spherical array of pulsed plasma guns to create a spherically imploding shell of very high velocity, high momentum flux plasma. This approach requires development of plasma jet accelerators capable of achieving velocities of 50-200 km/s with very precise timing and density profiles, and with high total mass and density. Low-Z plasma jets would require the higher velocities, whereas very dense high-Z plasma shells could achieve the goal at velocities of only 50-100 km/s. In this report, we describe our work to develop the pulsed plasma gun technology needed for an experimental scientific exploration of the PJMIF concept, and also for the other applications mentioned earlier. The initial goal of a few hundred of hydrogen at 200 km/s was eventually replaced with accelerating 8000 μg of argon or xenon to 50 km/s for the Plasma Liner Experiment (PLX) at Los Alamos National Laboratory (LANL). Initial work used existing computational and analytical tools to develop and refine a specific plasma gun concept having a novel tapered coaxial electromagnetic accelerator contour with an array of symmetric ablative plasma injectors. The profile is designed to suppress the main barrier to success in coaxial guns, namely the blow-by instability in which the arc slips past and outruns the bulk of the plasma mass. Efforts to begin developing a set of annular non-ablative plasma injectors for the coaxial gun, in order to accelerate pure gases, resulted in development of linear parallel-plate MiniRailguns that turned out to work well as plasma guns in their own right and we subsequently chose them for an initial plasma liner experiment on the PLX facility at LANL. This choice was mainly driven by cost and schedule for that particular experiment, while longer term goals still projected use of coaxial guns for reactor-relevant applications for reasons of better symmetry, lower impurities, more compact plasma jet formation, and higher gun efficiency. Our efforts have focused mainly on 1) developing various plasma injection systems for both coax and linear railguns and ensuring they work reliably with the accelerator section, 2) developing a suite of plasma and gun diagnostics, 3) performing computational modeling to design and refine the plasma guns, 4) establishing a research facility dedicated to plasma gun development, and finally, 5) developing plasma guns and associated pulse power systems capable of achieving these goals and installing and testing the first two gun sets on the PLX facility at LANL. During the second funding cycle for this program, HyperV joined in a collaborative effort with LANL, the University of Alabama at Huntsville, and the University of New Mexico to perform a plasma liner experiment (PLX) to investigate the physics and technology of forming spherically imploding plasma liners. HyperV’s tasks focused on developing the plasma guns and associated pulse power systems required for the 30 gun experiment at LANL. Unfortunately, funding for the entire PLX collaborative project was terminated after only two years of the four year project due to program funding realignments which necessitated recompeting the project in midstream. Despite the loss of funding, HyperV installed two Mark1 guns and pulsed power systems on PLX, and jet characterization and merging experiments were subsequently successfully performed at LANL by the PLX Team. In parallel with those PLX experiments, HyperV continued its efforts to develop a plasma gun capable of meeting the PLX goal of 8 mg of argon at 50 km/s. HyperV was ultimately successful in this effort, demonstrating 10.8 mg at 52.8 km/s and 7.5 mg at 62.4 km/s with the Mark2 MiniRailgun.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Zonal flow triggers the L-H transition in the Experimental Advanced Superconducting Tokamak

The kinetic energy transfer between shear flows and the ambient turbulence is investigated in the Experimental Advanced Superconducting Tokamak during the L-H transition. As the rate of energy transfer from the turbulence into the shear flow becomes comparable to the energy input rate into the turbulence, the transition into the H-mode occurs. As the observed behavior exhibits several predicted features of zonal flows, the results show the key role that zonal flows play in mediating the transition into H-mode.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Edge transport studies in the edge and scrape-off layer of the National Spherical Torus Experiment with Langmuir probes

Transport and turbulence profiles were directly evaluated using probes for the first time in the edge and scrape-off layer (SOL) of NSTX [Ono et al., Nucl. Fusion 40, 557 (2000)] in low (L) and high (H) confinement, low power (P{sub in}∼ 1.3 MW), beam-heated, lower single-null discharges. Radial turbulent particle fluxes peak near the last closed flux surface (LCFS) at ≈4×10{sup 21} s{sup −1} in L-mode and are suppressed to ≈0.2×10{sup 21} s{sup −1} in H mode (80%–90% lower) mostly due to a reduction in density fluctuation amplitude and of the phase between density and radial velocity fluctuations. The radial particle fluxes are consistent with particle inventory based on SOLPS fluid modeling. A strong intermittent component is identified. Hot, dense plasma filaments 4–10 cm in diameter, appear first ∼2 cm inside the LCFS at a rate of ∼1×10{sup 21} s{sup −1} and leave that region with radial speeds of ∼3–5 km/s, decaying as they travel through the SOL, while voids travel inward toward the core. Profiles of normalized fluctuations feature levels of 10% inside LCFS to ∼150% at the LCFS and SOL. Once properly normalized, the intermittency in NSTX falls in similar electrostatic instability regimes as seen in other devices. The L-H transition causes a drop in the intermittent filaments velocity, amplitude and number in the SOL, resulting in reduced outward transport away from the edge and a less dense SOL.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗