Recently published scenarios for fully non-inductive startup and operation on the National Spherical
Torus eXperiment Upgrade (NSTX-U) (Menard et al 2012 Nucl. Fusion 52 083015) show Electron
Cyclotron Resonance Heating (ECRH) as an important component in preparing a target plasma for
efficient High Harmonic Fast Wave and Neutral Beam heating. The modeling of the propagation and
absorption of EC waves in the evolving plasma is required to define the most effective window of
operation, and to optimize the launcher geometry for maximal heating and current drive during this
window. Here, we extend a previous optimization of O1-mode ECRH on NSTX-U to account for the
full time-dependent performance of the ECRH using simulations performed with TRANSP. We find
that the evolution of the density profile has a prominent role in the optimization by defining the time
window of operation, which in certain cases may be a more important metric to compare launcher
performance than the average power absorption. This feature cannot be captured by analysis on static
profiles, and should be accounted for when optimizing ECRH on any device that operates near the
cutoff density. Additionally, the utility of the electron Bernstein wave (EBW) in driving current and
generating closed flux surfaces in the early startup phase has been demonstrated on a number of
devices. Using standalone GENRAY simulations, we find that efficient EBW current drive is
possible on NSTX-U if the injection angle is shifted below the midplane and aimed towards the top
half of the vacuum vessel. However, collisional damping of the EBW is projected to be significant, in
some cases accounting for up to 97% of the absorbed EBW power
The National Spherical Torus Experiment Upgrade (NSTX-U) will advance the physics basis required for achieving steady-state, high-beta, and high-confinement conditions in a tokamak by accessing high toroidal field (1 T) and plasma current (1.0 - 2.0 MA) in a low aspect ratio geometry (A = 1.6 - 1.8) with flexible auxiliary heating systems (12 MW NBI, 6 MW HHFW). This paper describes progress in the development of L- and H-mode discharge scenarios and the commissioning of operational tools in the first ten weeks of operation that enable the scientific mission of NSTX-U. Vacuum field calculations completed prior to operations supported the rapid development and optimization of inductive breakdown at different values of ohmic solenoid current. The toroidal magnetic field (B_T0 = 0.65 T) exceeded the maximum values achieved on NSTX and novel long-pulse L-mode discharges with regular sawtooth activity exceeded the longest pulses produced on NSTX (tpulse > 1.8s). The increased flux of the central solenoid facilitated the development of stationary L-mode discharges over a range of density and plasma current (Ip). H-mode discharges achieved similar levels of stored energy, confinement (H98y,2 > 1) and stability (beta_N/beta_N-nowall > 1) compared to NSTX discharges for Ip < 1 MA. High-performance H-mode scenarios require an L-H transition early in the Ip ramp-up phase in order to obtain low internal inductance (li) throughout the discharge, which is conducive to maintaining vertical stability at high elongation (kappa > 2.2) and achieving long periods of MHD quiescent operations. The rapid progress in developing L- and H-mode scenarios in support of the scientific program was enabled by advances in real-time plasma control, efficient error field identification and correction, effective conditioning of the graphite wall and excellent diagnostic availability.
Kim, E.-W.; Bertelli, N.; Johnson, J.R.; Valeo, E.; Hosea, J.; Perkins, R.
We illustrate the capabilities of a recently developed two-dimensional full wave code (FW2D) in space and tokamak plasmas by adopting various values of density, magnetic field configuration and strength as well as boundary shape. As example, we first showed fast compressional wave propagation in the inner magnetosphere is dramatically modified by a plasmaspheric plume at Earth's magnetosphere. The results show that wave energy is trapped in the plume showing a leaky eigenmode-like structure with plume, which is similar to the detected magnetosonic waves. We also performed simulations of high harmonic fast waves in the scrape-off layer (SOL) plasmas of the National Spherical Torus eXperiment (NSTX)/NSTX-Upgrade. Comparison the results with previous full-wave simulations show that although the FW2D code uses a cold plasma approximation, the electric field and the fraction of the power losses in the SOL plasmas show excellent consistency and agreement with the previous full wave simulations performed by the AORSA code.
Kinetic modification of ideal stability theory from stabilizing resonances of mode-particle interaction has had success in explaining resistive wall mode (RWM) stability limits in tokamaks. With the goal of real-time stability forecasting, a reduced kinetic stability model has been implemented in the new Disruption Event Characterization and Forecasting (DECAF) code, which has been written to analyze disruptions in tokamaks. The reduced model incorporates parameterized models for ideal limits on beta, a ratio of plasma pressure to magnetic pressure, which are shown to be in good agreement with DCON code calculations. Increased beta between these ideal limits causes a shift in the unstable region of delta W_K space, where delta W_K is the change in potential energy due to kinetic effects that is solved for by the reduced model, such that it is possible for plasmas to be unstable at intermediate beta but stable at higher beta. Gaussian functions for delta W_K are defined as functions of E cross B frequency and collisionality, with parameters reflecting the experience of the National Spherical Torus Experiment (NSTX). The reduced model was tested on a database of discharges from NSTX and experimentally stable and unstable discharges were separated noticeably on a stability map in E cross B frequency, collisionality space. The reduced model only failed to predict an unstable RWM in 15.6% of cases with an experimentally unstable RWM and performed well on predicting stability for experimentally stable discharges as well.
To effectuate near real-time feedback control of ideal MHD instabilities in a tokamak geometry, a rapid solution for stability analysis is a prerequisite. Toward this end, we reformulate the δW stability method with a Hamilton-Jacobi theory, elucidating analytical and numerical features of the generic tokamak ideal MHD stability problem. The plasma response matrix is demonstrated to be the solution of an ideal MHD matrix Riccati differential equation (MRDE). Since Riccati equations are prevalent in the control theory literature, such a shift in perspective brings to bear a range of numerical methods that are well-suited to the robust, fast solution of control problems. We discuss the usefulness of Riccati techniques in solving the stiff ODEs often encountered in ideal MHD stability analyses-—for example, in tokamak edge and stellarator physics. We then demonstrate the applicability of such methods to an existing 2D ideal MHD stability code—DCON—enabling its parallel operation in near real-time. Output is shown to match with high accuracy, and with wall-clock time ≪ 1s. Such speed may help enable active feedback ideal MHD stability control, especially in tokamak plasmas whose ideal MHD equilibria evolve with inductive timescale τ > 1s-—as in ITER.
Halo currents generated during unmitigated tokamak disruptions are known to develop rotating asymmetric features that are of great concern to ITER because they can dynamically amplify the mechanical stresses on the machine. This paper presents a multi-machine analysis of these phenomena. More specifically, data from C-Mod, NSTX, ASDEX Upgrade, DIII-D, and JET are used to develop empirical scalings of three key quantities: (1) the machine-specific minimum current quench time, tauCQ; (2) the halo current rotation duration, trot; and (3) the average halo current rotation frequency, <fh>. These data reveal that the normalized rotation duration, trot/tauCQ, and the average rotation velocity, <vh>, are surprisingly consistent from machine to machine. Furthermore, comparisons between carbon and metal wall machines show that metal walls have minimal impact on the behavior of rotating halo currents. Finally, upon projecting to ITER, the empirical scalings indicate that substantial halo current rotation above <fh> = 20 Hz is to be expected. More importantly, depending on the projected value of tauCQ in ITER, substantial rotation could also occur in the resonant frequency range of 6-20 Hz. As such, the possibility of damaging halo current rotation during unmitigated disruptions in ITER cannot be ruled out.
Skinner, C.H.; Bedoya, F.; Scotti, F.; Allain, J.P.; Blanchard, W.; Cai, D.; Jaworski, M.; Koel, B.E.
Boronization has been effective in reducing plasma impurities and enabling access to higher density, higher confinement plasmas in many magnetic fusion devices. The National Spherical Torus eXperiment, NSTX, has recently undergone a major upgrade to NSTX-U in order to develop the physics basis for a ST-based Fusion Nuclear Science Facility (FNSF) with capability for double the toroidal field, plasma current, and NBI heating power and increased pulse duration from 1–1.5 s to 5–8 s. A new deuterated tri-methyl boron conditioning system was implemented together with a novel surface analysis diagnostic. We report on the spatial distribution of the boron deposition versus discharge pressure, gas injection and electrode location. The oxygen concentration of the plasma facing surface was measured by in-vacuo XPS and increased both with plasma exposure and with exposure to trace residual gases. This increase correlated with the rise of oxygen emission from the plasma.
Leading resistive wall mode (RWM) stability codes MARS-K [Y. Liu, et al., Phys. Plasmas 15, 112503 (2008)] and MISK [B. Hu, et al., Phys. Plasmas 12, 057301 (2005)] have been previously benchmarked. The benchmarking has now been extended to include additional physics, and used to project the stability of ITER in a realistic operating space. Due to ITER's relatively low plasma rotation and collisionality, collisions and non-resonance rotational effects were both found to have little impact on stability, and these non-resonance rotational effects also will not self-consistently affect the ITER RWM eigenfunction. Resonances between thermal ions and electrons and the expected level of ITER toroidal rotation were found to be important to stability, as were alpha particles, which are not in rotational resonance. MISK calculations show that without alpha particles, ITER is projected to be unstable to the RWM, but the expected level of alphas is calculated to provide a sufficient level of stability.
Measuring free-surface, liquid-metal flow velocity is challenging to do in a reliable and accurate manner. This paper presents a non-invasive, easily-calibrated method of measuring the surface velocities of open-channel liquid-metal flows using an IR camera. Unlike other spatially-limited methods, this IR camera particle tracking technique provides full field-of-view data that can be used to better understand open-channel flows and determine surface boundary conditions. This method could be implemented and automated for a wide range of liquid-metal experiments, even if they operate at high-temperatures or within strong magnetic fields.
Townsend avalanche theory is employed to model and interpret plasma initiation
in NSTX by Ohmic heating and coaxial helicity injection (CHI). The model is
informed by spatially resolved vacuum calculations of electric field and
magnetic field line connection length in the poloidal cross-section. The model
is shown to explain observations of Ohmic startup including the duration and
location of breakdown. Adapting the model to discharges initiated by CHI offers
insight into the causes of upper divertor (absorber) arcs in cases where the
discharge fails to initiate in the lower divertor gap. Finally, upper and lower
limits are established for vessel gas fill based on requirements for breakdown
and radiation. It is predicted that CHI experiments on NSTX-U should be
able to use as much as four times the amount of prefill gas employed in CHI
experiments in NSTX. This should provide greater flexibility for plasma
start-up, as the injector flux is projected to be increased in NSTX-U.