A design point is presented here for a prototype fusion neutron source for waste transmutation (10(17) - 10(19) n/s), based on the adiabatic compression of a compact torus (spheromak). The design utilizes the CORSICA (2D equilibrium) and NIMROD (3D time-dependent MHD) codes as well as analytic modeling with target parameters R-initial = 0.5 m, R-final = 0.167 m, T-initial = 0.4 keV, T-final = 4 keV, n(initial) = 2 x 10(20) m(-3) and n(final) = 50 x 10(20) m(-3), with radial convergence of C = 3. 3D time-dependent simulations of spheromak compression agree well with analytic models for adiabatic compression, if the run-in time tau(compress) < tau(E). Knowing tau(compress) required, we design coils and passive structure (with CORSICA) to ensure stability; then design the capacitor bank needed to both form the target plasma and drive coils. We specify target parameters for the compression in terms of plasma beta, formation efficiency and energy confinement.

The Facility for Laboratory Reconnection Experiments (FLARE) is an intermediate laboratory experiment currently under construction at Princeton University by a consortium of five universities and two Department of Energy (DoE) national laboratories, located at the Princeton Plasma Physics Laboratory (PPPL). The goal of FLARE is to provide experimental accesses to new regimes of the magnetic reconnection process and related phenomena directly relevant to heliophysics, astrophysics, and fusion plasmas. The device comprises a vacuum chamber and 9 coils sets that are independently programmable to provide the poloidal and toroidal magnetic fields required to form plasma and study the effects of magnetic reconnection. Each of these 9 coil sets requires a separate pulsed power system, it is the design of the power systems that is reported here. The 9 separate pulsed power systems combine to produce over 5.5MJ of energy to the experiment and each presented their own unique challenges. The most energetic power system is a 3.4MJ, 19.2mF capacitor bank charged to 20kV that provides the guide field, with a rise time of approximately 12ms it delivers an average peak current of 40kA over 5.3ms to 12 coils wired in series. The poloidal field coils consist of two separate coilsets each requiring 540kA peak current which is produced by two 20kV, 2.64mF capacitor banks. The design of the two driver coilsets each charged to 60kV will also be presented.

DIII-D experiments yield the first proof-of-principle results in feedback control of the proximity to the resistive wall mode (RWM) stability boundary using an active MHD spectroscopic stability measurement and neutral beam injection heating. In contrast to calculations of the stability of reconstructed equilibria, the spectroscopic measurement is independent of the assumed RWM stability model. The real-time implementation enables the control system to react to unforeseen changes in plasma parameters and hence stability limits. In the experimentally accessed regime, near but below the ideal-MHD no-wall limit for the n = 1 external kink instability, the control dynamics are described by a linear model that depends on the plasma stored energy. This model is used to aid in optimizing feedback gain settings.

New understanding and improved parameters have been achieved on the Helicity Injected Torus with Steady Inductive helicity injection current drive (HIT-SI) experiment. The experiment has a bowtie-shaped spheromak confinement region with two helicity injectors. The inductive injectors are 180° segments of a small, oval cross section toroidal pinch. Spheromaks with currents up to 38 kA and current amplification of 2 have been achieved with only 6 MW of injector power. The Taylor-state model is shown to agree with HIT-SI surface and internal magnetic profile measurements. Helicity balance predicts the peak magnitude of toroidal spheromak current and the threshold for spheromak formation. The model also accurately predicts the division of the applied loop voltage between the injector and spheromak regions. Single injector operation shows that the two injectors have opposing, preferred spheromak current directions. An electron locking relaxation model is consistent with the preferred direction, with ion Doppler data and with bolometric data. Results from higher frequency operation are given. The impact of the new understanding on the future direction of the HIT programme is discussed.

MHD simulations of rapid shutdown scenarios by massive particle injection in DIII-D, Alcator C-Mod and ITER are performed in order to study runaway electron (RE) transport during mitigated disruptions. The simulations include a RE confinement model using drift-orbit calculations for test particles. A comparison of limited and diverted plasma shapes is studied in DIII-D simulations, and improved confinement in the limited shape is found due to both spatial localization and reduced toroidal spectrum in the nonlinear MHD activity. C-Mod simulations compare shutdown scenarios in which impurity (Ar) fuelling is concentrated in the edge versus the core, and the confinement of REs in the core is maintained until the onset of the m = 1/n = 1 mode, which is delayed in the case of edge deposition, relative to core deposition. But, the overall RE loss fraction is 100% regardless of Ar fuelling profile. A comparison of simulations across the three devices points to a trend of increased RE confinement with increasing device size, wherein all REs are lost in C-Mod, all are confined in ITER, and a partial loss is observed in DIII-D. This trend is related to a reduction in the fluctuating field amplitude near the plasma edge during the thermal-quench-induced MHD activity. The result bodes poorly for RE mitigation strategies in ITER that rely on MHD deconfinement of REs.

Nonlinear MHD simulations of rapid shutdowns (or disruptions) in three tokamaks — DIII-D, Alcator C-Mod, and ITER — are performed with the NIMROD code. The simulation are initiated by depositing a large amount of neutral Ar in the plasma to initiate a radiative thermal quench, which modifies the current profile and leads to the onset of MHD instability. During the simulations, drift-orbits for a trace population of runaway electrons are integrated, in order to study the interaction between disruption-induced MHD and runaway electron confinement. DIII-D simulations show significantly greater runaway electron transport in diverted plasma disruptions that in limited plasma disruptions. In C-Mod, an edge-peaked [massive gas injection (MGI)-like] fueling profile is found to delay the loss of core confined runaway electrons relative to core-peaked (pellet-like) profile, although all runaway electrons are eventually lost in both cases. A comparison of results across all three devices suggests a strong reduction in runaway electron transport with increasing machine size, which is an unfavorable result for ITER.

Summary form only given. The Los Alamos National Laboratory (LANL) collaboration with Air Force Research Laboratory (AFRL) collaboration is close to a physics demonstration of compressional heating in a Magneto Inertial Fusion (MIF) plasma target. These first Magnetized Target Fusion (MTF) experiments will use solid aluminum flux compressor shells. The experimental high density Field Reversed Configuration (FRC) can be made to translate fast enough so that FRC lifetime is not an issue. We show some initial translation data from the Los Alamos FRC experiment FRXL that characterize the translated target plasma. We have taken advantage of the LANL experience so that a near duplicate of FRXL has come up in several months. The solid liner MTF is only one of several magnetized, pulsed MIF fusion schemes that are being pursued. We outline the present status of MTF including target formation, translation to a trapping region, and compression results.

Summary form only given. We present and overview the experimental high density Field Reversed Configurationi (FRC) approach for application to a physics demonstration of magnetized target fusion (MTF). This MT target plasma continues to be developed at the Los Alamos FRC experiment FRXL. The first translated FRXL FRC data will be shown, where the translation speeds exceed 15cm/usec, which yields a translation time substantially shorter than the FRC lifetimes. The conical theta coil is expected to generate toroidal magnetic field and helicity and increase stability and lifetime. The implications of the present data for MTF experiments will be discussed, along with the hardware, diagnostics, and pre-compression plasma formation and trapping experiments.

We apply an adiabatic evolution model to predict the behavior of a field reversed configuration (FRC) during decompression and translation, as well as during boundary compression. Semi-empirical scaling laws, which were developed and benchmarked primarily for collisionless FRCs, are expected to remain valid even for the collisional regime of FRX-L experiment. We use this approach to outline the design implications for FRX-L, the high density translated FRC experiment at Los Alamos National Laboratory. A conical theta coil is used to accelerate the FRC to the largest practical velocity so it can enter a mirror bounded compression region, where it must be a suitable target for a magnetized target fusion (MTF) implosion. FRX-L provides the physics basis for the integrated MTF plasma compression experiment at the Shiva-Star pulsed power facility at Kirtland Air Force Research Laboratory, where the FRC will be compressed inside a flux conserving cylindrical shell.

The Helicity Injected Torus with Steady Inductive Helicity Injection (HIT-SI) consists of an axisymmetric flux conserver and two half-torus magnetic helicity injectors, mounted on either side of the axisymmetric confinement region (Jarboe et al., 2006 , Phys. Rev. Lett., 97 , 115003). Current and flux are driven sinusoidally with time in each injector, injecting both power and magnetic helicity into the HIT-SI device, with the goal of forming and sustaining a spheromak in the confinement region. Recent HIT-SI results include formation of discharges with toroidal spheromak current 1.5 times the injector current amplitude, development of a Taylor-state model for the magnetic fields in HIT-SI discharges, and direct measurement of the portion of the induced injector electric field that drives current in the confinement region.