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Neutral beam injection (NBI) will be one of the main sources of heating and non-inductive current drive in ITER. Due to high level of injected power the beam induced heat loads present a potential threat to the integrity of the first wall of the device, particularly in the presence of non-axisymmetric perturbations of the magnetic field. Neutral beam injection can also destabilize Alfven eigenmodes and energetic particle modes, and act as a source of plasma rotation. Therefore, reliable and accurate simulation of NBI is important for making predictions for ITER, as well as for any other current or future fusion device. This paper introduces a new beamlet-based neutral beam ionization model called BBNBI. It takes into account the fine structure of the injector, follows the injected neutrals until ionization, and generates a source ensemble of ionized NBI test particles for slowing down calculations. BBNBI can be used as a stand-alone model but together with the particle following code ASCOT it forms a complete and sophisticated tool for simulating neutral beam injection. The test particle ensembles from BBNBI are found to agree well with those produced by PENCIL for JET, and those produced by NUBEAM both for JET and ASDEX Upgrade plasmas. The first comprehensive comparisons of beam slowing down profiles of interest from BBNBI + ASCOT with results from PENCIL and NUBEAM/TRANSP, for both JET and AUG, are presented. It is shown that, for an axisymmetric plasma, BBNBI + ASCOT and NUBEAM agree remarkably well. Together with earlier 3D studies, these results further validate using BBNBI + ASCOT also for studying phenomena that require particle following in a truly three-dimensional geometry.

The National Spherical Torus Experiment (NSTX) is currently being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam (NB) for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-Upgrade to achieve the research goals critical to a Fusion Nuclear Science Facility. These include producing stable, 100% non-inductive operation in high-performance plasmas, assessing plasma–material interface (PMI) solutions to handle the high heat loads expected in the next-step devices and exploring the unique spherical torus (ST) parameter regimes to advance predictive capability. Non-inductive operation and current profile control in NSTX-U will be facilitated by co-axial helicity injection (CHI) as well as radio frequency (RF) and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet–Parker theory. Full-wave AORSA simulations show that RF power losses in the scrape-off layer (SOL) increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. Toroidal Alfven eigenmode avalanches and higher frequency Alfven eigenmodes can affect NB-driven current through energy loss and redistribution of fast ions. The inclusion of rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing ideal wall and resistive wall modes. Neutral beams and neoclassical toroidal viscosity generated from applied 3D fields can be used as actuators to produce rotation profiles optimized for global stability. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade for disruption mitigation. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat flux handling. Simulations indicate that snowflake and impurity seeded radiative divertors are candidates for heat flux mitigation in NSTX-U. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium–oxygen affinity, which increases deuterium pumping and reduces recycling. In situ and test-stand experiments of lithiated graphite and molybdenum indicate temperature-enhanced sputtering, although that test-stand studies also show the potential for heat flux reduction through lithium vapour shielding. Non-linear gyro kinetic simulations have indicated that ion transport can be enhanced by a shear-flow instability, and that non-local effects are necessary to explain the observed rapid changes in plasma turbulence. Predictive simulations have shown agreement between a microtearing-based reduced transport model and the measured electron temperatures in a microtearing unstable regime. Two Alfven eigenmode-driven fast ion transport models have been developed and successfully benchmarked against NSTX data. Upgrade construction is moving on schedule with initial physics research operation of NSTX-U planned for mid-2015.

A conceptual, theoretical assessment of using the toroidal fusion reactor, tokamak, for deep space interplanetary and interstellar missions is presented. Toroidal thermonuclear fusion reactors, such as tokamaks and stellarators, are unique for space propulsion, allowing a design with the magnetic configuration localized inside the toroidal magnetic field coils. Plasma energetic ions, including charged fusion products, can escape such closed configuration at certain conditions as a result of vertical drift in the toroidal rippled magnetic field. Escaping particles can be used for direct propulsion (because toroidal drift is directed one way vertically) or to create and heat externally confined plasma, so that the latter can be used for propulsion. In contrast to other fusion concepts proposed for space propulsion, this concept utilizes the natural drift motion of charged particles out of the closed magnetic field configuration. Also, using deuterium-tritium (D-T) plasma is a novel way to use fusion neutrons with the energy of 14 MeV for direct propulsion (neutron rocket) for out of solar system missions. A special design of the blanket of the reactor allows neutrons to escape the device in a preferable direction. This provides a direct (partial) conversion of fusion energy into thrust.

Alfvenic instabilities (AEs) are well known as a potential cause of enhanced fast ion transport in fusion devices. Given a specific plasma scenario, quantitative predictions of (i) expected unstable AE spectrum and (ii) resulting fast ion transport are required to prevent or mitigate the AE-induced degradation in fusion performance. Reduced models are becoming an attractive tool to analyze existing scenarios as well as for scenario prediction in time-dependent simulations. In this work, a neutral beam heated NSTX discharge is used as reference to illustrate the potential of a reduced fast ion transport model, known as kick model, that has been recently implemented for interpretive and predictive analysis within the framework of the time-dependent tokamak transport code TRANSP. Predictive capabilities for AE stability and saturation amplitude are first assessed, based on given thermal plasma profiles only. Predictions are then compared to experimental results, and the interpretive capabilities of the model further discussed. Overall, the reduced model captures the main properties of the instabilities and associated effects on the fast ion population. Additional information from the actual experiment enables further tuning of the model's parameters to achieve a close match with measurements.

At the power levels required for significant heating and current drive in magnetically-confined toroidal plasma, modification of the particle distribution function from a Maxwellian shape is likely (Stix 1975 Nucl. Fusion 15 737), with consequent changes in wave propagation and in the location and amount of absorption. In order to study these effects computationally, both the finite-Larmor-radius and the high-harmonic fast wave (HHFW), versions of the full-wave, hot-plasma toroidal simulation code TORIC (Brambilla 1999 Plasma Phys. Control. Fusion 41 1 and Brambilla 2002 Plasma Phys. Control. Fusion 44 2423), have been extended to allow the prescription of arbitrary velocity distributions of the form . For hydrogen (H) minority heating of a deuterium (D) plasma with anisotropic Maxwellian H distributions, the fractional H absorption varies significantly with changes in parallel temperature but is essentially independent of perpendicular temperature. On the other hand, for HHFW regime with anisotropic Maxwellian fast ion distribution, the fractional beam ion absorption varies mainly with changes in the perpendicular temperature. The evaluation of the wave-field and power absorption, through the full wave solver, with the ion distribution function provided by either a Monte-Carlo particle and Fokker–Planck codes is also examined for Alcator C-Mod and NSTX plasmas. Non-Maxwellian effects generally tend to increase the absorption with respect to the equivalent Maxwellian distribution.

Toroidicity-induced Alfven eigenmode (TAE) stability in the National Spherical Torus Experiment (NSTX) [S. M. Kaye, M. Ono, Y.-K. M. Peng et al., Fusion Technol. 36, 16 (1999)] is analyzed using the improved NOVA-K code [N. N. Gorelenkov, C. Z. Cheng, and G. Y. Fu, Phys. Plasmas 6, 2802 (1999)], which includes finite orbit width and Larmor radius effects and is able to predict the saturation amplitude for the mode using the quasilinear theory. A broad spectrum of unstable global TAEs with different toroidal mode numbers is predicted. Due to the strong poloidal field and the presence of the magnetic well in NSTX, better particle confinement in the presence of TAEs in comparison with tokamaks is illustrated making use of the ORBIT code [R. B. White and M. S. Chance, Phys. Fluids 27, 2455 (1984)].

Tokamak plasmas can feature a large population of energetic particles (EP) from neutral beam injection or fusion reactions. In turn, energetic particles can drive instabilities, which affect the driving EP population leading to a distortion of the original EP distribution function and of quantities that depend on it. The latter include, for example, neutral beam (NB) current drive and plasma heating through EP thermalization. Those effects must be taken into account to enable reliable and quantitative simulations of discharges for present devices as well as predictions for future burning plasmas. Reduced models for EP transport are emerging as an effective tool for long time-scale integrated simulations of tokamak plasmas, possibly including the effects of instabilities on EP dynamics. Available models differ in how EP distribution properties are modified by instabilities, e.g. in terms of gradients in real or phase space. It is therefore crucial to assess to what extent different assumptions in the transport models affect predicted quantities such as EP profile, energy distribution, NB driven current and energy/momentum transfer to the thermal populations. A newly developed kick model, which includes modifications of the EP distribution by instabilities in both real and velocity space, is used in this work to investigate these issues. Coupled to TRANSP simulations, the kick model is used to analyze NB-heated NSTX and DIII-D discharges featuring unstable Alfvén eigenmodes (AEs). Results show that instabilities can strongly affect the EP distribution function, and modifications propagate to macroscopic quantities such as NB-driven current profile and NB power transferred to the thermal plasma species. Those important aspects are only qualitatively captured by simpler fast ion transport models that are based on radial diffusion of energetic ions only.

Neutral beam injection (NBI) is one of the primary tools foreseen for heating, current drive (CD) and q-profile control in future fusion reactors such as ITER and a Fusion Nuclear Science Facility. However, fast ions from NBI may also provide the drive for energetic particle-driven instabilities (e.g. Alfvenic modes (AEs)), which in turn redistribute fast ions in both space and energy, thus hampering the control capabilities and overall efficiency of NB-driven current. Based on experiments on the NSTX tokamak (M. Ono et al 2000 Nucl. Fusion 40 557), the effects of AEs and other low-frequency magneto-hydrodynamic instabilities on NB-CD efficiency are investigated. A new fast ion transport model, which accounts for particle transport in phase space as required for resonant AE perturbations, is utilized to obtain consistent simulations of NB-CD through the tokamak transport code TRANSP. It is found that instabilities do indeed reduce the NB-driven current density over most of the plasma radius by up to ~50%. Moreover, the details of the current profile evolution are sensitive to the specific model used to mimic the interaction between NB ions and instabilities. Implications for fast ion transport modeling in integrated tokamak simulations are briefly discussed.

Fast ion transport models currently implemented in the tokamak transport code TRANSP (Hawryluk 1980 Physics of Plasmas Close to Thermonuclear Conditions (Brussels: CEC)) are not capturing important aspects of the physics associated with resonant transport caused by instabilities such as toroidal Alfven eigenmodes (TAEs). This work describes the implementation of a fast ion transport model consistent with the basic mechanisms of resonant mode–particle interaction. The model is formulated in terms of a probability distribution function for the particle's steps in phase space, which is consistent with the Monte Carlo approach used in TRANSP. The proposed model is based on the analysis of the fast ion response to TAE modes through the ORBIT code (White and Chance 1984 Phys. Fluids 27 2455), but it can be generalized to higher frequency modes (e.g. compressional and global Alfven eigenmodes) and to other numerical codes or theories.

Energetic ion transport due to microturbulence is investigated in magnetohydrodynamic-quiescent plasmas by way of neutral beam injection in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)]. A range of on-axis and off-axis beam injection scenarios are employed to vary relevant parameters such as the character of the background microturbulence and the value of E b / T e, where Eb is the energetic ion energy and Te the electron temperature. In all cases, it is found that any transport enhancement due to microturbulence is too small to observe experimentally. These transport effects are modeled using numerical and analytic expectations that calculate the energetic ion diffusivity due to microturbulence. It is determined that energetic ion transport due to coherent fluctuations (e.g., Alfvén eigenmodes) is a considerably larger effect and should therefore be considered more important for ITER. © 2013 AIP Publishing LLC.

Fast beam-ion losses were studied in DIII-D in the presence of a scaled mock-up of two test blanket modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot, predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot, which is predicted to be different among the various codes. © 2011 IAEA, Vienna.

The sawtooth instability is known for inducing transport and loss of energetic particles (EPs), and for generating seed magnetic islands that can trigger tearing modes. Both effects degrade the overall plasma performance. Several theories and numerical models have been previously developed to quantify the expected EP transport caused by sawteeth, with various degrees of sophistication to differentiate the response of EPs at different energies and on different orbits (e.g. passing vs. trapped), although the analysis is frequently limited to a single time slice during a tokamak discharge. This work describes the development and initial benchmark of a framework that enables a reduced model for EP transport by sawteeth retaining the full EP phase-space information. The model, implemented in the ORBIT hamiltonian particle-following code, can be used either as a standalone post-processor taking input data from codes such as TRANSP, or as a pre-processor to compute transport coefficients that can be fed back to TRANSP for time-dependent simulations including the effects of sawteeth on EPs. The advantage of the latter approach is that the evolution of the EP distribution can be simulated quantitatively for sawtoothing discharges, thus enabling a more accurate modeling of sources, sinks and overall transport properties of EP and thermal plasma species for comprehensive physics studies that require detailed information of the fast-ion distribution function and its evolution over time.