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Disruption mitigation in tokamaks by impurity injection aims to reduce the heat load and mechanical forces and to collisionally suppress runaway electrons. Rapid injection of sufficient mass, high ...

We present an innovative idea to use hyper-velocity (>30 km/s) high-density (>1017 cm−3) plasma jets of D-T/H and C60-fullerene for magneto-inertial fusion (MIF), high energy density laboratory plasma (HEDLP), and disruption mitigation in magnetic fusion plasma devices. The mass (~1–2 g) of sublimated C60 and hydrogen (or D-T fuel) produced in a pulsed power source is ionized and accelerated as a plasma slug in a coaxial plasma accelerator. For MIF/HEDLP we propose to create a magnetized plasma target by injecting two high-Mach number high-density jets with fuel (D-T) and liner (C60/C) structure along the axis of a pulsed magnetic mirror. The magnetized target fusion (MTF) plasma created by head-on collision and stagnation of jets is compressed radially by a metallic liner (Z-pinch) and axially by the C60/C liner. For disruption mitigation, the C60 plasma jets were shown to be able to provide the required impurity mass (J Fusion Energy 27:6, 2008).

High-velocity hydrogen plasma jets demonstrated deep penetration into magnetically confined tokamak plasma during fueling experiments. Disruption mitigation with plasma jets requires much larger mass (∼0.3 g for DIII-D and ∼30 g for ITER) to be delivered into the plasma core. We present a concept of using hyper-velocity C60 fullerene-dusty plasma jets for disruption mitigation of magnetically confined plasmas. The C60-fullerenes can form a compact plasma slug which can be accelerated to hyper-velocity (∼50 km/s) allowing for deep penetration into the tokamak target plasma. We evaluate the principle advantages, means to create it using a pulsed source, the atomic processes during acceleration to hyper-velocity and subsequent transport, and the interaction with the ambient target tokamak plasma during penetration.

A set of twelve coils for stability control has recently been installed inside the DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] vacuum vessel, offering faster time response and a wider range of applied mode spectra than the previous external coils. Stabilization of the n=1 ideal kink mode is crucial to many high beta, steady-state tokamak scenarios. A resistive wall converts the kink to a slowly growing resistive wall mode (RWM). With feedback-controlled error field correction, rotational stabilization of the RWM has been sustained for more than 2.5 s. Using the internal coils, the required correction field is smaller than with the external coils, consistent with a better match to the mode spectrum of the error field. Initial experiments in direct feedback control have stabilized the RWMs at higher beta and lower rotation than could be achieved by the external coils in similar plasmas, in qualitative agreement with numerical modeling. The new coils have also allowed wall stabilization in plasmas with...

Recent DIII?D experiments show that ideal kink-modes can be stabilized at high beta by a resistive wall, with sufficient plasma rotation. However, the resonant response to static magnetic field asymmetries by a marginally stable resistive wall mode can lead to strong damping of the rotation. Careful reduction of such asymmetries has allowed plasmas with beta well above the ideal MHD no-wall limit, and approaching the ideal-wall limit, to be sustained for durations exceeding 1?s. Feedback control can improve plasma stability by direct stabilization of the resistive wall mode or by reducing magnetic field asymmetry. Assisted by plasma rotation, direct feedback control of resistive wall modes with growth rates more than five times faster than the characteristic wall time has been observed. These results open a new regime of tokamak operation above the free-boundary stability limit, accessible by a combination of plasma rotation and feedback control.

We report on the application of well-balanced Ross-filter systems for the diagnostics of x-ray emission from a Z-pinch plasma. The composition and thicknesses of the filter layers were so selected to yield the intensities of neon H- and He-like emission lines separately with relatively good accuracy. The systems provide convenient absolute and time-dependent measurements of the emission intensities, and they are particularly useful for relatively low source-light intensities. The applicability of the systems is examined with the aid of time-dependent modeling of the stagnating plasma, using collisional-radiative and radiation transport calculations. The data and modeling are used to yield information on time evolution of the plasma density and temperature at stagnation.

Results from an array of theoretical and computational tools developed to treat the instabilities of most interest for high performance tokamak discharges are described. The theory and experimental diagnostic capabilities have now been developed to the point where detailed predictions can be productively tested so that competing effects can be isolated and either eliminated or confirmed. The linear MHD stability predictions using high quality discharge equilibrium reconstructions are tested against the observations for the principal limiting phenomena in DIII-D: L mode negative central shear (NCS) disruptions, H mode NCS edge instabilities, and tearing and resistive wall modes (RWMs) in long pulse discharges. In the case of predominantly ideal plasma MHD instabilities, agreement between the code predictions and experimentally observed stability limits and thresholds can now be obtained to within several per cent, and the predicted fluctuations and growth rates to within the estimated experimental errors. Edge instabilities can be explained by a new model for edge localized modes as predominantly ideal instabilities with low to intermediate toroidal mode number. Accurate ideal calculations are critical to demonstrating RWM stabilization by plasma rotation, and the ideal eigenfunctions provide a good representation of the RWM structure when the plasma rotation slows. Ideal eigenfunctions can then be used to predict stabilization using active feedback. For non-ideal modes, the agreement in some cases is promising. Δ' calculations, for example, indicate that some discharges are linearly unstable to classical tearing modes, consistent with the observed growth of islands in those discharges. Nevertheless, there is still a great deal of improvement required before the non-ideal predictive capability can routinely approach levels similar to those for the ideal comparisons.

Active feedback control in the DIII-D tokamak has fully stabilized the current-driven ideal kink resistive-wall mode (RWM). While complete stabilization is known to require both low frequency error field correction (EFC) and high frequency feedback, unambiguous identification has been made about the distinctive role of each in a fully feedback-stabilized discharge. Specifically, the role of direct RWM feedback, which nullifies the RWM perturbation in a time scale faster than the mode growth time, cannot be replaced by low frequency EFC, which minimizes the lack of axisymmetry of external magnetic fields.

We present the progress on the development of a new idea of using high-Mach number high-density composite plasma jets from coaxial plasma guns for disruption mitigation in tokamak1 and magneto-inertial fusion2 (MIF). The key element of the idea is the solid state pulsed power source with TiH 2 (or TiDT) grains and C 60 micron size powder3. Very fast injection of the molecular gas mixture provided by hydrogen release and sublimation of C 60 into the plasma gun is achieved by a special filter grid with supersonic Laval nozzles. The estimations based on the physical models of TiH2 grains heating, C 60 powder sublimation, molecular gas injection, mass separation, and plasma slug acceleration will be detailed. For disruption mitigation, our calculations show that the plasma gun is able to provide the required impurity mass1 (∼1–2 g) and the ram pressure to penetrate the tokamak hot plasma and to overcome the confining magnetic field pressure. Core tokamak plasma penetration can be achieved and impurity mass delivered in less than 1 ms, as required by ITER tokamak. The magnetized target fusion (MTF) plasma for MIF is created by injecting two high-Mach number (M≫5) high-density (≫1017 cm−3) plasma jets composed of fuel (D-T) and “pusher” (C 60 /C) along the axis of a pulsed magnetic (∼1–2 T) mirror into a metallic cylindrical liner. The high-density (∼1018 cm−3) cylindrical MTF created by head-on collision and stagnation in the magnetic field is compressed radially by the Z-pinch of the liner and prevented to expand axially by the incoming C 60 /C end-plugs. We estimated that, due to the much longer MTF axial dimension (∼30 cm) as compared to other inertial confinement fusion plasmas, the electron thermal conduction time to the C 60 /C end-plugs is longer than liner implosion time.

The resistive wall mode (RWM) and neoclassical tearing mode (NTM) have been simultaneously suppressed in the DIII-D for durations of over 2 s at beta values 20% above the no-wall limit with modest electron cyclotron current drive and very low plasma rotation. The achieved plasma rotation was significantly lower than reported previously. However, in this regime where stable operation is obtained, it is not unconditionally guaranteed. Various MHD activities, such as edge localized modes (ELMs) and fishbones, begin to couple to the RWM branch near the no-wall limit; feedback has been useful in improving the discharge stability to such perturbations. Simultaneous operation of slow dynamic error field correction and fast feedback suppressed the pile-up of ELM-induced RWM at a series of ELM events. This result implies that successful feedback operation requires not only direct feedback against unstable RWM but also careful control of MHD-induced RWM aftermath, which is the dynamical response to a small-uncorrected error field near the no-wall beta limit. These findings are extremely useful in defining the challenge of control of the RWM and NTM in the unexplored physics territory of burning plasmas in ITER.

Numerical simulation results by the GEM and MCBC codes are presented, along with a comparison with experiments for beam capture dynamics and parameter studies of charge state distribution (CSD) of electron cyclotron resonance charge breeder ion sources. First, steady state plasma profiles are presented by GEM with respect to key experimental parameters such as rf power and gas pressure. As rf power increases, electron density increases by a small amount and electron energy by a large amount. The central electrostatic potential dip also increased. Next, MCBC is used to trace injected beam ions to obtain beam capture profiles. Using the captured ion profiles, GEM obtains a CSD of beam ions. As backscattering can be significant, capturing the ions near the center of the device enhances the CSD. The effect of rf power on the beam CSD is mainly due to different steady states plasmas. Example cases are presented assuming that the beam ions are small enough not to affect the plasma.

FAR-TECH, Inc. has been building up a suite of comprehensive numerical tools for end-to-end electron cyclotron resonance ion source (ECRIS) modeling. They consist of the Monte Carlo beam capture (MCBC) code [1, 2], the generalized ECRIS modeling (GEM) code [3], and the ion extraction (IonEx) code [4]. The MCBC code simulates beam slowing down dynamics in a plasma due to Coulomb collisions, and includes ionization due to hot electrons and charge exchange. GEM models ECR plasmas by fluid ions and bounce averaged electrons. MCBC provides ion source profiles to GEM, which in turn provides ion flux profiles to IonEx. IonEx accurately models the plasma meniscus formed during ion extraction using an innovative, meshfree, particle-in-cloud-of-points technique. Extensions of GEM to two dimensions (2D) and IonEx to three dimensions (3D), as well as experimental validation are underway.

Semiconductor detectors can provide measurements of the X-ray emission spectra from Z-pinch plasmas with spectral, time and spatial resolution simultaneously. PIN detectors working in the current regime (up to 1 A), fast enough (1 nsec rise time), and small enough (1 to 3 mm/sup 2/ active area) are suitable, provided the X-ray spectral resolution is achieved by some element external to the detector before the X-ray flux is reaching its sensitive area. Ross (or balanced) filters are especially useful for this X-ray filtration system. A Ross filter system was designed and built for the investigation of the Ne IX and Ne X line emission. The experiments were carried out on the moderate density and temperature Ne plasma (14 mm length and approximately

A new formula for the magnetic piston model, which explicitly describes how the momentum imparted to the ions by the magnetic pressure depends not only on the ion mass but also on the ion charge, as well as, on the plasma electron and ion temperatures, is derived following Rosenbluth's classical particle-field self-consistent plane approximation analytic calculation. The formula presented in this paper has implications in explaining the experimentally observed separation of the ions of different species and charges by the magnetic field penetrating the plasma and specularly reflecting them.

The injection of Ar into the region of the DIII-D divertor is a promising technique for energy dissipation (through radiation and collisions) and consequently for reduction of the heat load on the plates. An important problem related to this technique, is the inherent poisoning of the core plasma by migrating Ar. The Ar core contamination seems also to improve the thermal transport in an advanced operating mode of the tokamak. It is therefore of great importance to measure the evolution of the impurity concentration profile within the core plasma. This goal could be achieved by using the Ross filter method in conjunction with the existing x-ray diagnostics on DIII-D. A basic Ross filter system consists of two identical detectors placed behind two different x-ray absorbing foils looking at the same plasma volume. The foils are made of different elements or compounds with adjacent or nearly adjacent atomic numbers. Their accurate thickness causes the x-ray transmission curves of the two foils to be effectiv...

Techniques to reduce the heat flux to the divertor plates in tokamak power plants and the consequent erosion of, and subsequent damage to the divertor target plates include the injection of impurities such as argon, that can dissipate the energy (through radiative or collisional processes) before it reaches the target plates. An important issue with this type of scheme is poisoning of the plasma core by the impurities introduced in the divertor region. Subsequently, there is a desire to measure the profiles of the injected impurities in the core. X-ray Ross filters with an effective narrow band pass centered on the argon Kα line at 3.2 keV, have been installed on two of the existing x-ray arrays on DIII–D in order to help determine the argon concentration profiles. Emissivity profiles of the Kα lines and the emissivity profiles for the argon enhanced continuum can be inferred from the inverted filtered x-ray brightness signals if Te, ne, and Ar18+ profiles are known. The MISTReference 1 code is used to co...