The National Ignition Facility at Lawrence Livermore National Laboratory was formally dedicated in May 2009. The hohlraum energetics campaign with all 192 beams began shortly thereafter and ran until early December 2009. These experiments explored hohlraum-operating regimes in preparation for experiments with layered cryogenic targets. The hohlraum energetic series culminated with an experiment that irradiated an ignition scale hohlraum with 1 MJ. The results demonstrated the ability to produce a 285 eV radiation environment in an ignition scale hohlraum while meeting ignition requirements for symmetry, backscatter and hot electron production. Complementary scaling experiments indicate that with ~1.3 MJ, the capsule drive temperature will reach 300 eV, the point design temperature for the first ignition campaign. Preparation for cryo-layered implosions included installation of a variety of nuclear diagnostics, cryogenic layering target positioner, advanced optics and facility mod...

The resonant magnetic perturbation (RMP) fields, including the plasma response, are computed within a linear, full toroidal, single-fluid resistive magnetohydrodynamic (MHD) model, and under realistic plasma conditions for MAST and ITER. The response field is found to be considerably reduced, compared with the vacuum field produced by the magnetic perturbation coils. This field reduction relies strongly on the screening effect from the toroidal plasma rotation. Computations also quantify three-dimensional (3D) distortions of the plasma surface, caused by RMP fields. A correlation is found between the computed mode structures, the plasma surface displacement and the observed density pump-out effect in MAST experiments. Generally, the density pump-out tends to occur when the surface displacement peaks near the X-points.

Since the end of 2005 most of the plasma–wall interaction experiments on FTU have been focused on the possible use of liquid lithium as the plasma facing material. Liquid lithium limiter is an active method to deposit, during the plasma discharge, a lithium film on the walls with prolonged beneficial effects. Reliable operation with very clean plasmas, very low wall particle recycling, spontaneous peaking of the density profile for line-averaged density values have been obtained. These results have allowed us to extend the density limit to the highest value so far obtained ( at I _p = 0.7 MA and B _T = 7.1 T, q _a = 5.0, by gas puffing o...

Localized current injection near the outboard midplane is used to form 0.1 MA plasma discharges with no induction supplied from a central solenoid in the ultra-low aspect ratio Pegasus Toroidal Experiment. The discharges are initiated by driving open-field-line currents that perturb the vacuum magnetic field such that the magnetic topology transitions to a tokamak-like configuration. The plasma is subsequently driven via helicity injection from the edge current sources and poloidal field induction. Intermittent n = 1 MHD activity is observed during periods of strong edge current drive and each event leads to a rapid inward expansion of the plasma volume and a drop in the plasma inductance. The plasmas are sufficiently turbulent such that the equilibrium approaches the lowest energy state described by Taylor relaxation theory. In agreement with that theory, the maximum I _p scales with ( I _TF I _inj/ w) ^1/2, whe...

A series of carefully designed experiments on DIII-D have taken advantage of a broad set of turbulence and profile diagnostics to rigorously test gyrokinetic turbulence simulations. In this paper the goals, tools and experiments performed in these validation studies are reviewed and specific examples presented. It is found that predictions of transport and fluctuation levels in the mid-core region (0.4 < ρ < 0.75) are in better agreement with experiment than those in the outer region (ρ ≥ 0.75) where edge coupling effects may become increasingly important and multiscale simulations may also be necessary. Validation studies such as these are crucial in developing confidence in a first-principles based predictive capability for ITER.

Generation of initial closed flux surfaces by electron cyclotron heating (ECH) at a conventional aspect ratio of R/ a ~ 3 is investigated on the Low Aspect ratio Torus Experiment (LATE) device and further tested on the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) tokamak. In both experiments, a plasma current is initiated and increased by ECH under steady external fields composed of a toroidal field and a weak vertical field. In the LATE experiments, a movable inboard limiter is used to change the aspect ratio of plasmas up to R/ a ~ 3. The results show that the formation of closed flux surfaces is still possible up to R/ a ~ 3 while higher decay indices of vertical field are required as the aspect ratio increases. Similar current start-up discharges are performed on JT-60U, and a plasma current is initiated and increased up to 20 kA under a vertical field having a high decay index.

Along with an expanded evaluation of the equilibrium operating space of the Korea Superconducting Tokamak Advanced Research, KSTAR, experimental equilibria of the most recent plasma discharges were reconstructed using the EFIT code. In near-circular plasmas created in 2009, equilibria reached a stored energy of 54 kJ with a maximum plasma current of 0.34 MA. Highly shaped plasmas with near double-null configuration in 2010 achieved H-mode with clear edge localized mode (ELM) activity, and transiently reached a stored energy of up to 257 kJ, elongation of 1.96 and normalized beta of 1.3. The plasma current reached 0.7 MA. Projecting active and passive stabilization of global MHD instabilities for operation above the ideal no-wall beta limit using the designed control hardware was also considered. Kinetic modification of the ideal MHD n = 1 stability criterion was computed by the MISK code on KSTAR theoretical equilibria with a plasma current of 2 MA, internal inductance of ...

The Omega Laser Facility is used to study inertial confinement fusion (ICF) concepts. This paper describes progress in direct-drive central hot-spot (CHS) ICF, shock ignition (SI) and fast ignition (FI) since the 2008 IAEA FEC conference. CHS cryogenic deuterium–tritium (DT) target implosions on OMEGA have produced the highest DT areal densities yet measured in ICF implosions (~300 mg cm ⁻²). Integrated FI experiments have shown a significant increase in neutron yield caused by an appropriately timed high-intensity, high-energy laser pulse.

W and W-alloys are among the primary candidate materials for plasma-facing components in the design of fusion reactors, particularly in high-heat-flux regions such as the divertor. Under neutron irradiation W undergoes transmutation to its near-neighbours in the periodic table. Additionally He and H are particles emitted from certain neutron-induced reactions, and this is particularly significant in fusion research since the presence of helium in a material can cause both swelling and a strong increase in brittleness. This paper presents the results of inventory burn-up calculations on pure W and gives quantitative estimates for He production rates in both a fusion-reactor environment and under conditions expected in the ITER experimental device. Transmutation reactions in possible alloying elements (Re, Ta, Ti and V), which could be used to reduce the brittleness of pure W, are also considered. Additionally, for comparison, the transmutation of other fusion-relevant materials, i...