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Fast neutrons can perform two functions that are critical to the nuclear power industry: breed fissile material from fertile material and burn minor actinides in spent fuel. The fusion reaction in a fusion reactor between deuterium and tritium nuclei yields an alpha particle and a neutron with 14.5-MeV energy. These neutrons can easily convert fertile material (such as 238U and 232Th) into fissile material (such as 239Pu and 233U) and destroy minor actinides (Pu, Np, Am, and Cm) present in the spent nuclear fuel and significantly reduce its radiotoxicity. However, each neutron generated implies one tritium consumed and tritium is generated by nuclear reactions between a neutron and lithium nuclei. Therefore, neutron economy becomes a crucial issue and an adequate neutronic modeling of the neutron source becomes critical. In this article, the Automatic System for TRansport Analysis (ASTRA) and Monte Carlo N-Particle (MCNP) codes are used in conjunction to characterize the neutron flux coming out of a low-aspect-ratio tokamak machine. ASTRA is first used to determine temperature and density profiles within the plasma volume and MCNP is used to evaluate the neutron flux. Since the simulation of an extended neutron source with complex geometry such as the confined plasma inside the tokamak is difficult, steps are taken to simplify the geometry by defining two simple sources, a cylindrical wall and a disk, which should have similar neutron emission distributions to the volumetric source. This reduction is done using a preprocessing Monte Carlo algorithm to establish probability density functions of neutron emission for the cylindrical wall as a function of the height and for the disk as a function of the radius, based on the emission characteristics of the extended source. The results that compare the neutron flux leaving a volume for four different configurations of neutron sources (isotropic source at the geometrical origin, ring source at the plasma geometrical center, cylindrical wall source, and distributed multipoint source) are presented. It is found that the processing of the two simple sources has a higher computational cost in terms of both processing time and random number consumption.

Fast neutron irradiation has the capability of transforming fertile material, such as 238U or 232Th, into fissile material that can be used to manufacture fuel for conventional thermal light-water fission reactors. This would help both expand and extend the lifetime of nuclear power, which is currently constrained to natural uranium resources, which could only last between 50 and 150 years depending on the growth of the nuclear reactor fleet. In this paper, a design for a fast neutron source based on a low-aspect ratio tokamak is evaluated, based on two figures of merit: the tritium breeding ratio (TBR), a measure of tritium self-sufficiency, and the reactor support ratio (RSR), related to the speed of fissile material breeding. The system has neutron multiplying regions to ensure there are enough neutrons to breed both tritium and fissile material; it also has tritium breeding regions that contain 6Li which, when bombarded with neutrons, produce tritium. The MCNP code is used to perform neutronics simulations and obtain values for energy-resolved neutron fluxes for the relevant regions of the system, and the ORIGEN code is used to perform kinetic calculations for nuclear reactions between the neutrons and the materials within the regions. Simulations show that for the case of a spherical tokamak with aspect ratio of 1.8 at 120 MW of neutron power, the tritium breeding ratio can have values ranging 0.85 and 1.25, and with that power level the system can bring 336 assemblies to 5% 233U enrichment in irradiation times between 24 and 65 months. The values depend on both the selection of materials and key geometry parameters, such as the thickness of container walls, as well as the volumes for both neutron multiplying and tritium breeding regions. FLiBe was found to give the best performance as neutron multiplier, since it contributes to the tritium self-sufficiency as well; from the tritium breeding perspective, lithium oxide was found to be the best choice.

The next generation fusion machines are likely to face enormous heat exhaust problems. In addition to summarizing major issues and physical processes connected with these problems, we discuss how advanced divertors, obtained by modifying the local geometry, may yield workable solutions. We also point out that: (1) the initial interpretation of recent experiments show that the advantages, predicted, for instance, for the X-divertor (in particular, being able to run a detached operation at high pedestal pressure) correlate very well with observations, and (2) the X-D geometry could be implemented on ITER (and DEMOS) respecting all the relevant constraints. A roadmap for future research efforts is proposed.

Nuclear fuel cycles based on thorium are gaining close attention due to its higher availability and more homogenous geographical distribution. Thorium cycles, likely to be less problematic with regard to waste generation and weapons proliferation, will extend the availability of nuclear fuel by hundreds/thousands of years. The principal Th cycle involves the transmutation of the fertile 232Th isotope into the fissile isotope 233U by means of neutron capture. In the present study, the coupled operation of a hybrid fission/fusion system and a standard thermal reactor is analyzed. Using the MCNP neutronic transport code, the behavior for 233U consumption/generation in both systems as a function of Th/U feed ratio to each reactor is analyzed. The useful composition range for the feed to the thermal reactor was found to be between 1.7% and 2.25%; within these range, breeder input enrichment can be found which causes the rate of consumption and generation of 233U to be identical. Under this condition, from the point of view of the fissile isotope, the cycle is closed, with no net generation or consumption. The cycle requires a 232Th input to compensate the amount spent on breeding the fissile material; for the range of interest, this varies between 0 and 35% of the total mass flow in the breeder leg of the cycle.

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th 232 –U 233 conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO 2 matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U 235 fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

The use of a fusion component testing facility to study and establish, during the ITER era, the remaining scientific and technical knowledge needed by fusion Demo is considered and described in thi...

The national high-power advanced torus experiment is a concept for a new facility to address the FESAC theme of ‘taming the plasma-material interface’. This concept exploits the compactness and excellent access provided by low aspect ratio to achieve a high ratio of exhaust power to major radius in order to study the integration of high-performance, long-pulse plasmas with a reactor-relevant high heat flux plasma boundary. Predictions of the scrape-off-layer plasma characteristics are presented, as calculated with the 2D edge modeling code SOLPS. Calculations in a variety of magnetic geometries indicate that very high levels of divertor heat flux can be expected, with peak values far in excess of the power handling capabilities of presently-used materials. Possible methods to reduce the heat flux to acceptable levels are discussed.

A fusion-assisted transmutation system for the destruction of transuranic nuclear waste is developed by combining a subcritical fusion–fission hybrid assembly uniquely equipped to burn the worst thermal nonfissile transuranic isotopes with a new fuel cycle that uses cheaper light water reactors for most of the transmutation. The center piece of this fuel cycle, the high power density compact fusion neutron source (100 MW, outer radius

Laboratory experiments using a plasma wind generated by laser-target interaction are developed to investigate the creation of a shock in front of the magnetosphere and the dynamo mechanism. Magnetic obstacles are placed in the plasma wind and measurements of the electron density gradients surrounding the obstacles are recorded. The experiments are analyzed with the methods used in theoretical simulation of the solar-wind-driven magnetosphere interactions. The proposed experiments are thought to be relevant to understanding the electron acceleration mechanisms at work in shock-driven magnetic dipole confined plasma surrounding compact magnetized stars and planets. Electron shock acceleration mechanisms are discussed in some detail.

In MBX, a small mirror machine with a radial electric field creates a rotating plasma that is expected to evolve, under certain conditions, into a self-organizing, detached toroidal plasma ring, a magnetofluid state. In the present stage of the experiment a low density plasma generated by microwaves (1 kW at 2.54 GHz) has been successfully rotated at supersonic speeds using a 1 kV–80 mF capacitor bank with currents ∼5 amps. Under these conditions the plasma presents high asymmetry in the current, plasma potential and consequently rotation with the voltage applied. A simple model is presented to account for these features.

An experimental simulation of planetary magnetospheres is being developed to investigate the formation of collisionless shocks and their effects. Two experimental situations are considered. In both, the solar wind is simulated by laser ablation plasmas. In one case, the “solar wind” flows across the magnetic field of a high-current discharge. In the other, a transverse magnetic field is embedded in the plasma flow, which interacts with a conductive obstacle. The ablation plasma is created using the “Tomcat” laser, currently emitting 5 J in a 6 ns pulse at 1 μm wavelength and irradiance above 1013 W/cm2. The “Zebra” z-pinch generator, with load current up to 1 MA and voltage up to 3.5 MV produces the magnetic fields. Hydrodynamic modeling is used to estimate the plasma parameters achievable at the front of the plasma flow and to optimize the experiment design. Particle-in-cell simulations reveal details of the interaction of the “solar wind” with an external magnetic field, including flow collimation and heating effects at the stopping point. Hybrid simulations show the formation of a bow shock at the interaction of a magnetized plasma flow with a conductor. The plasma density and the embedded field have characteristic spatial modulations in the shock region, with abrupt jumps and fine structure on the skin depth scale.

Recent experimental observations in tokamaks indicate enhanced convection of plasma blobs toward the main chamber wall. Potential implications of these observations for reactors are examined here. Two-dimensional plasma edge calculations are performed with UEDGE, including convective transport consistent with present experiments. This is coupled to a kinetic neutral calculation using the code NUT, to compute the hot neutral flux to the wall. The inclusion of convection increases sputtering of the wall by roughly an order of magnitude. For tungsten walls, erosion (neglecting re-deposition) is estimated to be ∼.6 mm year−1. The enhanced source of impurities for high Z walls requires an enhanced plasma screening factor to allow ignition. Low Z liquid materials enable acceptable plasma contamination with much lower screening factors. Rough estimates of dust generation from erosion rates with convection imply significant safety issues. Plasma transport via blobs can also strongly modify models of impurity screening and redeposition, and represents a potential feasibility issue in need of further research.

Due to the X-point singularity, the safety factor tends to infinity as approaching to the last closed flux surface. The numerical treatments of the near X-point behavior become challenging both for equilibrium and stability. The usual solution is to cut off a small fraction of edge region for system stability evaluation or simply use an up-down symmetric equilibrium without X-point as an approximation. In this work, we assess the sensitivity of this type of equilibrium treatments on the stability calculation. It is found that the system stability can depend strongly on the safety factor value (qa) at the edge after the cutting-off. When the edge safety factor value falls in the vicinity of a rational mode number (referred to as the resonant gap), the system becomes quite unstable due to the excitation of the peeling type modes. Instead, when the edge safety factor is outside the resonant gaps, the system is much more stable and the predominant modes become the usual external kink (or ballooning and infern...

The X-divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at 10–20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. However, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. The model also points to carbon radiation as the primary driver of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency for core operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.

The first high fidelity gyrokinetic simulations of the energy losses in the transport barriers of large tokamaks in pursuit of fusion gain are presented. These simulations calculate the turbulent energy losses with an extensive treatment of relevant physical effects—fully kinetic, non-linear, electromagnetic—inclusive of all major plasma species, and in equilibria with relevant shape and local bootstrap current for fusion-relevant cases. We find that large plasmas with a small normalized gyroradius lie in an unexpected regime of enhanced losses that can prevent the projected energy gain. Our simulations are qualitatively consistent with recent experiments on JET with an ITER-like wall. Interestingly and very importantly, the simulations predict parameter regimes of reduced transport that are quite fusion-favorable.

JET has been unable to recover historical confinement levels when operating with an ITER-like wall (ILW) due largely to the inaccessibility of high pedestal temperatures. Finding a path to overcome this challenge is of utmost importance for both a prospective JET DT campaign and for future ITER operation. Gyrokinetic simulations (using the Gene code) quantitatively capture experimental transport levels for a representative experimental discharge and qualitatively recover the major experimental trends. Microtearing turbulence is a major transport mechanisms for the low-temperature pedestals characteristic of unseeded JET-ILW discharges. At higher temperatures and/or lower , we identify electrostatic ITG transport of a type that is strongly shear-suppressed on smaller machines. Consistent with observations, this transport mechanism is strongly reduced by the presence of a low-Z impurity (e.g. carbon or nitrogen at the level of ), recovering the accessibility of high pedestal temperatures. Notably, simulations based on dimensionless scans recover historical scaling behavior except in the unique JET-ILW parameter regime where ITG turbulence becomes important. Our simulations also elucidate the observed degradation of confinement caused by gas puffing, emphasizing the important role of the density pedestal structure. This study maps out important regions of parameter space, providing insights that may point to optimal physical regimes that can enable the recovery of high pedestal temperatures on JET.

The diamagnetic drift effects on the low-$n$ magnetohydrodynamic instabilities at the high-mode (H-mode) pedestal are investigated in this paper with the inclusion of bootstrap current for equilibrium and rotation effects for stability, where $n$ is the toroidal mode number. The AEGIS (Adaptive EiGenfunction Independent Solutions) code [L. J. Zheng, M. T. Kotschenreuther, J. Comp. Phys. {\bf 211}, (2006)] is extended to include the diamagnetic drift effects. This can be viewed as the lowest order approximation of the finite Larmor radius effects in consideration of the pressure gradient steepness at the pedestal. The H-mode discharges at Jointed European Torus (JET) is reconstructed numerically using the VMEC code [P. Hirshman and J. C. Whitson, Phys. Fluids {\bf 26}, 3553 (1983)], with bootstrap current taken into account. Generally speaking, the diamagnetic drift effects are stabilizing. Our results show that the effectiveness of diamagnetic stabilization depends sensitively on the safe factor value ($q_s$) at the safety-factor reversal or plateau region. The diamagnetic stabilization are weaker, when $q_s$ is larger than an integer; while stronger, when $q_s$ is smaller or less larger than an integer. We also find that the diamagnetic drift effects also depend sensitively on the rotation direction. The diamagnetic stabilization in the co-rotation case is stronger than in the counter rotation case with respect to the ion diamagnetic drift direction., 29 pages, 13 figures

A new global optimization technique for designing stellarator coils has been developed and applied to the design of coils for the National Compact Stellarator Experiment. Using this technique coil sets were found with fewer coils and lower current densities than those obtained with traditional methods. A new coil design procedure which uses a genetic algorithm as the core optimization method is described and the resulting advanced coil designs presented.

Physics issues are discussed for compact stellarator configurations which achieve good confinement by the fact that the magnetic field modulus |B| in magnetic co-ordinates is dominated by poloidally symmetric components. Two distinct configuration types are considered: (1) those which achieve their drift optimization and rotational transform at low β and low bootstrap current by appropriate plasma shaping; and (2) those which have a greater reliance on plasma β and bootstrap currents for supplying the transform and obtaining quasi-poloidal symmetry. Stability analysis of the latter group of devices against ballooning, kink and vertical displacement modes has indicated that stable β values on the order of 15% are possible. The first class of devices is being considered for a low β near term experiment that could explore some of the confinement features of the high β configurations.

Experimental devices for the study of the physics of high beta (β 4%), low aspect ratio (A 4.5) stellarator plasmas require coils that will produce plasmas satisfying a set of physics goals, provide experimental flexibility and be practical to construct. In the course of designing a flexible coil set for the National Compact Stellarator Experiment, several innovations have been made that may be useful in future stellarator design efforts. These include: the use of singular value decomposition methods for obtaining families of smooth current potentials on distant coil winding surfaces from which low current density solutions may be identified; the use of a control matrix method for identifying which few of the many detailed elements of a stellarator boundary must be targeted if a coil set is to provide fields to control the essential physics of the plasma; the use of a genetic algorithm for choosing an optimal set of discrete coils from a continuum of potential contours; the evaluation of alternate coil topologies for balancing the trade-off between physics objectives and engineering constraints; the development of a new coil optimization code for designing modular coils and the identification of a `natural' basis for describing current sheet distributions.

Significant progress has been made in the development of new, compact stellarator devices that could test optimization principles for the design of a more attractive reactor. These are 3 and 4 field period low aspect ratio quasi-omnigenous stellarators based on an optimization method that aims at improved confinement, stability, ease of coil design, low aspect ratio and low bootstrap current. With respect to the analysis of these devices, the authors have developed improved measures of neoclassical transport, collisionless loss analyses of ICRF tail and α population confinement and methods for the rapid assessment of ballooning stability boundaries.

Key physics issues in the design of a high- quasi-axisymmetric stellarator configuration are discussed. The goal of the design study is a compact stellarator configuration with aspect ratio comparable to that of tokamaks and good transport and stability properties. Quasi- axisymmetry has been used to provide good drift trajectories. Ballooning stabilization has been accomplished by strong axisymmetric shaping, yielding a stellarator configuration whose core is in the second stability regime for ballooning modes. A combination of externally generated shear and non-axisymmetric corrugation of the plasma boundary provides stability to external kink modes even in the absence of a conducting wall. The resulting configuration is also found to be robustly stable to vertical modes, increasing the freedom to perform axisymmetric shaping. Stability to neoclassical tearing modes is conferred by a monotonically increasing profile. A gyrokinetic f code has been used to confirm the adequacy of the neoclassical confinement. Neutral beam losses have been evaluated with Monte Carlo codes.

Recent progress in the theoretical understanding and design of compact stellarators is described. Hybrid devices, which depart from canonical stellarators by deriving benefits from the bootstrap current which flows at finite beta, comprise a class of low aspect ratio A

A diagnostic neutral beam will be installed on Alcator C-Mod for measurement of ni, nz, Tz, vθ, vφ, and j. The beam will be used primarily for the study of thermal transport, Er in H-mode discharges, and to provide critical profile and turbulence data for tests of theoretically based models of turbulence and transport. The beam selected for this work was used on Texas Experimental Tokamak and is rated at 50 kV and 6 A of extracted current in hydrogen. It can produce beams of deuterium or helium as well. It is currently being reconfigured for installation on C-Mod. Capabilities for the intended measurements are described using simulations based on measured C-Mod profiles.

One of the perceived barriers for the sustainability of nuclear energy is the issue of the high-level waste in the form of spent fuel. This issue affects the public image of the industry as well as the rentability of a nuclear operation. Strategies aimed at the reduction and utilization of spent fuel are hence desirable to reduce the radiotoxicity of the waste and reduce the volume requirement for a geological repository. There are proven technologies that can extract U and Pu isotopes suitable to be burned on thermal reactors, but the minor actinides and the nonfissile isotopes of Pu represent a problem, since they are the main contributors to the radiotoxicity of spent fuel by being decay chain heads. Their early elimination significantly reduces the activity of the spent fuel over time. Fusion-fission hybrids are being considered as an alternative for actinide burning and transmutation taking advantage of the energetic neutrons generated in the fusion reactor, in order to burn actinides present in a fission blanket around the fusion neutron source. In the present paper, operating scenarios for the Compact Fusion Neutron Source (CFNS) fed by reprocessed PWR spent fuel are considered. The goal of the simulations is to optimize the combination of coolant, matrix and fuel composition such that the maximization of both the actinide burning and the energy multiplication in the system is achieved, with the fission blanket generating excess energy after feeding the fusion neutron source; the systems is also subject to the constrain of being a subcritical assembly for safety considerations. The neutronic behavior for different combinations of gas coolants (Li, CO2 and He) was modeled using the MCNPX code using the materials and the geometry already defined for CFNS, and using typical spent fuel minor actinides concentration and composition.

The first nonlinear gyrokinetic turbulence simulations that quantitatively reproduce experimental transport levels in an H-mode pedestal are reported. In the JET-ILW (ITER-like wall) pedestal, the bulk of the transport in the steep gradient region is caused by the turbulence driven by the microtearing mode (MTM). Kinetic ballooning modes are found to be in a second-stability regime. With contributions from the neoclassical and electron temperature gradient driven transport, the MTM mechanism reproduces, quantitatively, the experimental power balance across most of the pedestal.

Recent favorable results on START have caused renewed interest in low aspect ratio tokamaks. To design an economical next-step spherical tokamak to study confinement scaling and high beta plasmas, we have developed a transport scaling and device optimization code. This code OPT, benchmarked against START, includes 10 empirical confinement scaling laws and essential tokamak physics such as stability limits. Parameters are optimized separately for each scaling law and physics goal. Using OPT we find for R/a=1.2 to 2.0 one can achieve {beta}{sub N}=5 and [{beta}]=30% with just two neutral beams (P{sub NB} =0.75 MA, and R{sub 0}>=0.6 m. In contrast, if one insists on using the nominal device parameters, I{sub p}=1 MA and R{sub 0}=0.8 m, with each scaling law, achieving {beta}{sub N}=5 requires typically P{sub NB}>=7.5 MW. 8 refs., 2 figs., 3 tabs.

Monte Carlo simulation studies for ion orbit loss in limiter tokamaks show a poloidal asymmetry in ion loss arising from differences in ion orbit geometry. Since electron loss to the limiter is uniformly distributed because of its tiny orbit width, the nonuniform ion loss could cause a poloidal electric field that would tend to make the ion loss to the limiter more uniform. A simple analytical derivation of this poloidal electric field and a discussion of its effects on ion movement and transport are also presented.

Comparison of experiments with theory leads to a reasonably complete understanding of impurity rotation in Ohmic plasmas in a tokamak. In the experiments, the components of the impurity rotation velocity were derived from measurements of the Doppler shift of carbon spectral emission. Neoclassical theory for rotation was evaluated using measured plasma profiles and compared with experiment. The comparisons suggest that plasma toroidal rotation is damped (but not quenched) by momentum loss due to ion–neutral charge exchange and that the damping can now be accurately predicted.

Closed form Green’s functions are derived for convective–diffusive systems in cylindrical geometries. They are of the form e−αr 2I0(βr), where I0 is the modified Bessel function and α and β depend on time via e−t. These solutions can be used to model a variety of time‐dependent transport experiments, such as pulsed impurity transport, oscillating gas puff, perturbative rf heating, and sawtooth heat pulse propagation. A successful comparison of the authors’ profiles with pulsed impurity injection experiments on the Texas Experimental Tokamak (TEXT) [K. W. Gentle, Nucl. Technol./Fusion 1, 479 (1981)] is presented.

Modulation of the gas feed to a tokamak induces a density modulation and also a temperature modulation. Although the density change is fractionally small and the response is linear, the perturbed particle flux can exceed the equilibrium flux. The particle flux {Gamma} thus cannot be a simple homogeneous function of {del}{ital n}. The ratio of thermal to particle flux is quite different for the equilibrium and perturbations, precluding a simple linkage of particle and thermal transport. The modulation of the amplitude of density turbulence does not correlate with any of the modulated fluxes.

The authors presents a plasma density perturbation analysis which closely couples measurements of perturbations in density, temperature and turbulent fluctuation levels with a theoretical model in order to calculate particle fluxes. Non-linearities in the plasma model and the modulation of plasma parameters such as the electron temperature caused by the density perturbation are found to be significant. Some necessary details of the modification of the density fluctuation spectrum, and the plasma potential and Zeff profile remain to be measured. However, the available data allow steady state and perturbed particle fluxes in the low collisionality region of TEXT (5 cm < r < 20 cm) to be calculated using a simplified dissipative trapped electron model. These calculated fluxes do not agree with the measured fluxes

The authors present an analytical calculation of the neutral particle distribution and its effects on ion heat and momentum transport in three-dimensional plasmas with arbitrary temperature and density profiles. A general variational principle, taking advantage of the simplicity of the charge exchange (CX) operator, is derived to solve self-consistently the problem of neutral-plasma interaction. To facilitate an extremal solution, the short CX mean-free-path (λx) ordering is used. Furthermore, a non-variational, analytical solution providing a full set of transport coefficients is derived by making the realistic assumption that the product of the CX cross-section with relative velocity is constant. The effects of neutrals on plasma energy loss and rotation appear in simple, sensible forms. It is found that neutral viscosity dominates ion viscosity everywhere, and in the edge region by a large factor

We present a fast code for calculating the steady-state transport of neutral atoms in an axially symmetric background plasma. The primary source for the neutrals is not required to have any symmetry. Due to the small momentum transfer involved in charge exchange collisions, the secondary neutral atoms emerge with the local velocity distribution of the plasma ions. Some neutrals are lost due to ionization. The neutral transport is described by an integral equation for the neutral source. With a careful choice of a three-dimensional spatial grid, the small scale features of these equations can be integrated analytically to yield a set of algebraic equations that can be solved by only a few iterations. This results in a fast and compact algorithm which can be used as a subroutine in plasma simulation codes. Comparisons with other currently used codes and with experimental measurements show good agreement.

The carbon concentration in the central region of TFTR discharges with high power neutral beam heating has been measured by charge exchange recombination spectroscopy (CXRS) of the C5+ n = 3–4 transition in the extreme ultraviolet region. The carbon concentrations were deduced from absolute measurements of the line brightness using a calculation of the beam attenuation and the appropriate cascade corrected line excitation rates. As a result of the high ion temperatures (20–30 keV) in most of the discharges, the contribution of beam halo neutrals to the line brightness was significant and therefore had to be included in the modelling of the data. Carbon concentrations have been measured in discharges with plasma currents Ip in the range 1.0-1.6 MA and beam power in the range 2.6–30 MW, including a number of supershots. The results are in good agreement with carbon concentrations deduced from the visible bremsstrahlung Zeff and with metallic impurity concentrations measured by X-ray pulse height analysis, demonstrating the reliability of the atomic rates used in the beam attenuation and line excitation calculations. Carbon is the dominant impurity species in these discharges; the oxygen concentration measured via CXRS in a high beam power case was 0.0006 of ne, compared to 0.04 for carbon. Trends with plasma current and beam power in the carbon concentration and the inferred deuteron concentration are presented. The carbon concentration is independent of plasma current and decreases from 0.13 at 2.6 MW beam power to 0.04 at 30 MW, while the deuteron concentration increases from 0.25 to 0.75 over the same range of beam power. These changes are primarily the result of beam particle fuelling, as the carbon density did not vary significantly with beam power. The time evolutions of the carbon and deuteron concentrations during two high power beam pulses, one which exhibited a carbon bloom (a sudden influx of carbon due to local heating of the limiter) and one which did not, are compared. In both types of discharge, the carbon concentration decreases early in the beam pulse as a result of beam particle fuelling, and the carbon density rises slowly during the beam pulse until the start of the bloom. The electron density rise during the bloom is primarily due to the increase in the carbon density.

We investigate laboratory simulations of magnetic astrophysical jets using optical probing. We have observed differences in laser-produced plasmas with and without a magnetic field using interferometric measurements.

The MBX experiment was built during the grant period. Details can be found in publications about MBX experiment. We created a low density plasma in a mirror configuration and rotated it at supersonic speeds in the theta direction. Under these conditions the plasma presents a high asymmetry in the current, plasma potential and consequently rotation with the voltage applied. We developed a two fluid model to describe the measurements based on a magnetofluid states ansatz. We also observed bistable and multiple states predicted by the two fluid model.

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)-a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (>= 6.5 T, 1.5 MA), high power density facility (P/S similar to 1.5 MW m(-2)) will test innovative divertor ideas, including an 'X-point target divertor' concept, at the required performance parameters-reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region-while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magneticfield side-the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination-advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions-will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv: 1409.3540)) that makes use of high-temperature superconductor technology-a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

We present initial results from a new q diagnostic for TEXT‐Upgrade. This method is based on using a toroidal array of detectors to determine the plane in which beam‐injected neutrals are emitted after two charge exchange collisions. The potential advantages are low cost, full plasma accessibility, and good time resolution. Our initial series of experiments of TEXT‐U establish the feasibility of this technique.

It is found that the X-divertor (XD) configuration (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA), CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502) can be made with the conventional poloidal field (PF) coil set on ITER (Tomabechi et al and Team 1991 Nucl. Fusion 31 1135), where all PF coils are outside the TF coils. Starting from the standard divertor, a sequence of desirable XD configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. This opens the possibility that the XD could be tested and used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the super-X-divertor (Kotschenreuther et al 2007 Bull. Am. Phys. Soc. 53 11, Valanju et al 2009 Phys. Plasmas 16 5, Kotschenreuther et al 2010 Nucl. Fusion 50 035003, Valanju et al 2010 Fusion Eng. Des. 85 46) is not a feature of the XD geometry. In addition, we present an XD configuration for K-DEMO (Kim et al 2013 Fusion Eng. Des. 88 123) to demonstrate that it is also possible to attain the XD configuration in advanced tokamak reactors with all PF coils outside the TF coils. The results given here for the XD are far more encouraging than recent calculations by Lackner and Zohm (2012 Fusion Sci. Technol. 63 43) for the Snowflake (Ryutov 2007 Phys. Plasmas 14 064502, Ryutov et al 2008 Phys. Plasmas 15 092501), where the required high PF currents represent a major technological challenge. The magnetic field structure in the outboard divertor SOL (Kotschenreuther 2013 Phys. Plasmas 20 102507) in the recently created XD configurations reproduces what was presented in the earlier XD papers (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA) CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502). Consequently, the same advantages accrue, but no close-in PF coils are employed.

Relying on coil positions relative to the plasma, the “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the “proximity condition,” used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors.

We present the first treatment of the refraction of physical electromagnetic waves in newly developed negative index media (NIM), also known as left-handed media (LHM). The NIM dispersion relation implies that group fronts refract positively even when phase fronts refract negatively. This difference results in rapidly dispersing, very inhomogeneous waves. In fact, causality and finite signal speed always prevent negative wave signal (not phase) refraction. Earlier interpretations of phase refraction as ``negative light refraction'' and ``light focusing by plane slabs'' are therefore incorrect, and published NIM experiments can be explained without invoking negative signal refraction.

We use a diagnostic neutral beam, a scanning neutral particle energy analyzer, and Hα measurements to estimate the neutral density profile from the edge to the center of the TEXT tokamak. The diagnostic neutral beam provides a spatially localized density of neutrals. The plasma neutral density is measured from the ratio of the signals seen with and without the beam. The Hα measurements provide neutral profiles in the edge region and the global particle confinement time τp. The shapes of the observed neutral energy spectra put a lower bound on the central neutral density. Results are compared with estimates of NUT and general agreement is found. Error estimates are given.

Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust - the Scrape-Off Layer (SOL). A primary result of our analysis is the emergence of a physical "metric", the Divertor Index DI, that quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics - the Standard Divertor (SD, DI = 1), and two advanced geometries: the X-Divertor (XD, DI > 1) and the Snowflake (SFD, DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent NSTX and DIIID experiments are X-Divertors, not Snowflakes., From Institute for Fusion Studies, The University of Texas at Austin. 34 pages, 16 figures, 2 tables

Magnetohydrodynamic instabilities of high-mode (H-mode) pedestal are investigated in this paper with the inclusion of bootstrap current for equilibrium and rotation for stability. The jointed European torus-like equilibria of H-mode discharges are generated numerically using the VMEC code. It is found that, when the bootstrap current is taken into account, a safety-factor reversal or plateau can be generated near plasma edge. This confirms previous results of numerical equilibrium reconstructions using other types of codes. The n = 1 magnetohydrodynamic instabilities, where n is toroidal mode number, are investigated numerically in this type of equilibria using the AEGIS code. It is found that the infernal type harmonic can prevail at safety-factor reversal or plateau region. The toroidal plasma rotation effect with low Mach number is investigated. The numerical results show that the mode frequency is close to the rotation frequency at pedestal top, when the value of safety factor at plateau is slightly above a rational number. This mode frequency range seems to coincide with the experimentally observed frequencies of n = 1 edge harmonic oscillations (or outer modes) at the quiescent H-mode discharges.

Magnetized collimated plasma jets are created in the laboratory to extend our understanding of plasma jet acceleration and collimation mechanisms with particular connection to astrophysical jets. In this study, plasma collimated jets are formed from supersonic unmagnetized flows, mimicking a stellar wind, subject to currents and magnetohydrodynamic forces. It is found that an external poloidal magnetic field, like the ones found anchored to accretion disks, is essential to stabilize the jets against current-driven instabilities. The maximum jet length before instabilities develop is proportional to the field strength and the length threshold agrees well with Kruskal-Shafranov theory. The plasma evolution is modeled qualitatively using MHD theory of current-carrying flux tubes showing that jet acceleration and collimation arise as a result of electromagnetic forces.

A new magnetic geometry, the super X divertor (SXD), is invented to solve severe heat exhaust problems in high power density fusion plasmas. SXD divertor plates are moved to the largest major radii inside the TF coils, increasing the wetted area by 2–3 and the line length by 2–5. Two-dimensional fluid simulations with SOLPS (Schneider et al 2006 SOLPS 2-D edge calculation code Contrib. Plasma Phys. 46) show a several-fold decrease in divertor heat flux and plasma temperature at the plate. A small high power density tokamak using SXD is proposed, for either (1) useful fusion applications using conservative physics, such as a component test facility (CTF) or fission–fusion hybrid, or (2) to develop more advanced physics modes for a pure fusion reactor in an integrated fusion environment.

This volume contains a unique collection of papers contributed by experts with a long association with George Sudarshan in seven different areas of physics. Each paper recognizes the seminal contributions to physics made by George Sudarshan, and acknowledges their impact on their own research. They were presented at the conference Sudarshan: Seven Science Quests held at the University of Texas at Austin in November 2006. Participants spanned George's 5 decades of research and represented three of the 4 inhabited continents. Our only regret is that due to the vast breadth of George's quests, some topics (e.g. Quantum Field Theory), had to be left out. The conference was truly a celebration of Physics itself, due to the participation and contributions of a galaxy of stellar physicists who are leaders in their chosen fields. The focal point of the conference was to showcase George Sudarshan's breakthrough initiation of these seven areas of physics, and to celebrate his 75th birthday. Many of his former and current students, colleagues, friends and family gathered in Austin, Texas for 2 days and presented their research that was based on his prior and current works, and beautiful personal memories. Papers presented traced the history of the origin of the seven physics fields in which George's role as the originator/initiator has often been overlooked or misrepresented, and the status of these fields in current times. Even many of his colleagues who have worked with him on one or two of the quests were unaware of the impressive breadth and depth of his contributions to the other quests. We feel that the conference achieved its objective of creating a unified showcase of some of his major contributions in a single volume. Of course, knowing George, his quests are ongoing. We wish him many more years of productive pursuit of science. This meeting would not have been possible without the financial support of several institutions. We are deeply indebted to the Office of the Vice President of Research, the Physics Department, and the Center for Electromagnetic Materials and Devices of The University of Texas at Austin; and to MetaMaterials LLC, Austin,Texas. Finally, we would like to thank the participants, and particularly George's family, for their contribution to the wonderful atmosphere achieved during the Conference. We would like also to acknowledge the authors of the papers collected in the present volume, the members of the Scientific Committee for their guidance and support, and the referees for their generous work.