Jin Myung Park

Off-axis neutral beam current drive for advanced scenario development in DIII-D

Modification of the two existing DIII-D neutral beamlines is planned to allow vertical steering to provide off-axis neutral beam current drive (NBCD) peaked as far off-axis as half the plasma minor radius. New calculations for a downward-steered beam indicate strong current drive with good localization off-axis so long as the toroidal magnetic field, BT, and the plasma current, Ip, point in the same direction. This is due to good alignment of neutral beam injection (NBI) with the local pitch of the magnetic field lines. This model has been tested experimentally on DIII-D by injecting equatorially mounted NBs into reduced size plasmas that are vertically displaced with respect to the vessel midplane. The existence of off-axis NBCD is evident in the changes seen in sawtooth behaviour in the internal inductance. By shifting the plasma upwards or downwards, or by changing the sign of the toroidal field, off-axis NBCD profiles measured with motional Stark effect data and internal loop voltage show a difference in amplitude (40–45%) consistent with differences predicted by the changed NBI alignment with respect to the helicity of the magnetic field lines. The effects of NBI direction relative to field line helicity can be large even in ITER: off-axis NBCD can be increased by more than 30% if the BT direction is reversed. Modification of the DIII-D NB system will strongly support scenario development for ITER and future tokamaks as well as provide flexible scientific tools for understanding transport, energetic particles and heating and current drive.

Integrated modeling applications for tokamak experiments with OMFIT

One modeling framework for integrated tasks (OMFIT) is a comprehensive integrated modeling framework which has been developed to enable physics codes to interact in complicated workflows, and support scientists at all stages of the modeling cycle. The OMFIT development follows a unique bottom-up approach, where the framework design and capabilities organically evolve to support progressive integration of the components that are required to accomplish physics goals of increasing complexity. OMFIT provides a workflow for easily generating full kinetic equilibrium reconstructions that are constrained by magnetic and motional Stark effect measurements, and kinetic profile information that includes fast-ion pressure modeled by a transport code. It was found that magnetic measurements can be used to quantify the amount of anomalous fast-ion diffusion that is present in DIII-D discharges, and provide an estimate that is consistent with what would be needed for transport simulations to match the measured neutron rates. OMFIT was used to streamline edge-stability analyses, and evaluate the effect of resonant magnetic perturbation (RMP) on the pedestal stability, which have been found to be consistent with the experimental observations. The development of a five-dimensional numerical fluid model for estimating the effects of the interaction between magnetohydrodynamic (MHD) and microturbulence, and its systematic verification against analytic models was also supported by the framework. OMFIT was used for optimizing an innovative high-harmonic fast wave system proposed for DIII-D. For a parallel refractive index , the conditions for strong electron-Landau damping were found to be independent of launched and poloidal angle. OMFIT has been the platform of choice for developing a neural-network based approach to efficiently perform a non-linear multivariate regression of local transport fluxes as a function of local dimensionless parameters. Transport predictions for thousands of DIII-D discharges showed excellent agreement with the power balance calculations across the whole plasma radius and over a broad range of operating regimes. Concerning predictive transport simulations, the framework made possible the design and automation of a workflow that enables self-consistent predictions of kinetic profiles and the plasma equilibrium. It is found that the feedback between the transport fluxes and plasma equilibrium can significantly affect the kinetic profiles predictions. Such a rich set of results provide tangible evidence of how bottom-up approaches can potentially provide a fast track to integrated modeling solutions that are functional, cost-effective, and in sync with the research effort of the community.

Demonstration of ITER operational scenarios on DIII-D

The DIII-D programme has recently initiated an effort to provide suitably scaled experimental evaluations of four primary ITER operational scenarios. New and unique features of this work are that the plasmas incorporate essential features of the ITER scenarios and anticipated operating characteristics; e.g. the plasma cross-section, aspect ratio and value of I/aB of the DIII-D discharges match the ITER design, with size reduced by a factor of 3.7. Key aspects of all four scenarios, such as target values for βN and H98, have been replicated successfully on DIII-D, providing an improved and unified physics basis for transport and stability modelling, as well as for performance extrapolation to ITER. In all four scenarios, normalized performance equals or closely approaches that required to realize the physics and technology goals of ITER, and projections of the DIII-D discharges are consistent with ITER achieving its goals of ≥400 MW of fusion power production and Q ≥ 10. These studies also address many of the key physics issues related to the ITER design, including the L–H transition power threshold, the size of edge localized modes, pedestal parameter scaling, the impact of tearing modes on confinement and disruptivity, beta limits and the required capabilities of the plasma control system. An example of direct influence on the ITER design from this work is a modification of the physics requirements for the poloidal field coil set at 15 MA, based on observations that the inductance in the baseline scenario case evolves to a value that lies outside the original ITER specification.

Beam-ion confinement for different injection geometries

The DIII-D tokamak is equipped with neutral beam sources that inject in four different directions; in addition, the plasma can be moved up or down to compare off-axis with on-axis injection. Fast-ion data for eight different conditions have been obtained: co/counter, near-tangential/near-perpendicular and on-axis/off-axis. Neutron measurements during short beam pulses assess prompt and delayed losses under low-power conditions. As expected, co-injection has fewer losses than counter, tangential fewer than perpendicular and on-axis fewer than off-axis; the differences are greater at low current than at higher current. The helicity of the magnetic field has a weak effect on the overall confinement. Fast-ion Dα (FIDA) and neutron measurements diagnose the confinement at higher power. The basic trends are the same as in low-power plasmas but, even in plasmas without long wavelength Alfven modes or other MHD, discrepancies with theory are observed, especially in higher temperature plasmas. At modest temperature, two-dimensional images of the FIDA light are in good agreement with the simulations for both on-axis and off-axis injection. Discrepancies with theory are more pronounced at low fast-ion energy and at high plasma temperature, suggesting that fast-ion transport by microturbulence is responsible for the anomalies.

Non-inductive current drive and transport in high βN plasmas in JET

A route to stationary MHD stable operation at high ?N has been explored at the Joint European Torus (JET) by optimizing the current ramp-up, heating start time and the waveform of neutral beam injection (NBI) power. In these scenarios the current ramp-up has been accompanied by plasma pre-heat (or the NBI has been started before the current flat-top) and NBI power up to 22?MW has been applied during the current flat-top. In the discharges considered transient total ?N ? 3.3 and stationary (during high power phase) ?N ? 3 have been achieved by applying the feedback control of ?N with the NBI power in configurations with monotonic or flat core safety factor profile and without an internal transport barrier (ITB). The transport and current drive in this scenario is analysed here by using the TRANSP and ASTRA codes. The interpretative analysis performed with TRANSP shows that 50?70% of current is driven non-inductively; half of this current is due to the bootstrap current which has a broad profile since an ITB was deliberately avoided. The GLF23 transport model predicts the temperature profiles within a ?22% discrepancy with the measurements over the explored parameter space. Predictive simulations with this model show that the E ? B rotational shear plays an important role for thermal ion transport in this scenario, producing up to a 40% increase of the ion temperature. By applying transport and current drive models validated in self-consistent simulations of given reference scenarios in a wider parameter space, the requirements for fully non-inductive stationary operation at JET are estimated. It is shown that the strong stiffness of the temperature profiles predicted by the GLF23 model restricts the bootstrap current at larger heating power. In this situation full non-inductive operation without an ITB can be rather expensive strongly relying on the external non-inductive current drive sources.

Simulations of KSTAR high performance steady state operation scenarios

We report the results of predictive modelling of high performance steady state operation scenarios in KSTAR. Firstly, the capabilities of steady state operation are investigated with time-dependent simulations using a freeboundary plasma equilibrium evolution code coupled with transport calculations. Secondly, the reproducibility of high performance steady state operation scenarios developed in the DIII-D tokamak, of similar size to that of KSTAR, is investigated using the experimental data taken from DIII-D. Finally, the capability of ITER-relevant steady state operation is investigated in KSTAR. It is found that KSTAR is able to establish high performance steady state operation scenarios; βN above 3, H98(y, 2) up to 2.0, fBS up to 0.76 and fNI equals 1.0. In this work, a realistic density profile is newly introduced for predictive simulations by employing the scaling law of a density peaking factor. The influence of the current ramp-up scenario and the transport model is discussed with respect to the fusion performance and non-inductive current drive fraction in the transport simulations. As observed in the experiments, both the heating and the plasma current waveforms in the current ramp-up phase produce a strong effect on the q-profile, the fusion performance and also on the non-inductive current drive fraction in the current flattop phase. A criterion in terms of qmin is found to establish ITER-relevant steady state operation scenarios. This will provide a guideline for designing the current ramp-up phase in KSTAR. It is observed that the transport model also affects the predictive values of fusion performance as well as the non-inductive current drive fraction. The Weiland transport model predicts the highest fusion performance as well as non-inductive current drive fraction in KSTAR. In contrast, the GLF23 model exhibits the lowest ones. ITER-relevant advanced scenarios cannot be obtained with the GLF23 model in the conditions given in this work. Finally, ideal MHD stability is investigated for the ITER-relevant advanced scenarios in KSTAR. The methods and results presented in this paper are expected to contribute to improving the ITER and beyond ITER predictive simulations.

Prospects for Off-Axis Neutral Beam Current Drive in the DIII-D Tokamak

Abstract Off-axis neutral beam (NB) current drive (CD) (NBCD) has the potential to supply substantial off-axis CD for the demonstration steady-state, Advanced Tokamak scenarios. A modification of the two existing DIII-D NB beamlines is proposed to allow off-axis CD with NB injection (NBI) vertically steered to drive current as far off axis as half the plasma radius. The profile and magnitude of the driven current is calculated using the NUBEAM Monte Carlo module in the TRANSP and ONETWO transport codes. When the beam is injected in the same direction as the toroidal field, off-axis CD of [approximately equal to]45 kA/MW is calculated at normalized radius (square root of the toroidal flux), ρ = 0.5 with full-width at half-maximum of 0.45 in ρ. The dimensionless CD efficiency is comparable or somewhat better than that for electron cyclotron CD (ECCD) at the same location and plasma parameters. The efficiency stays nearly constant in going from on-axis to off-axis CD. The localization and magnitude of the off-axis NBCD are sensitive to the alignment of the NBI relative to the helical pitch of the magnetic field lines and thus to the direction of the toroidal field and plasma current. The driven current is still localized off axis for fast ion diffusivities up to 1 m2/s. The calculations show that the off-axis NBCD can supply much of the off-axis CD for the steady-state scenarios under consideration, leaving ECCD for fine-tuning of the current profile and real-time control.

An efficient transport solver for tokamak plasmas

Abstract A simple approach to efficiently solve a coupled set of 1-D diffusion-type transport equations with a stiff transport model for tokamak plasmas is presented based on the 4th order accurate Interpolated Differential Operator scheme along with a nonlinear iteration method derived from a root-finding algorithm. Numerical tests using the Trapped Gyro-Landau-Fluid model show that the presented high order method provides an accurate transport solution using a small number of grid points with robust nonlinear convergence.

Combined magnetic and kinetic control of advanced tokamak steady state scenarios based on semi-empirical modelling

This paper shows that semi-empirical data-driven models based on a two-time-scale approximation for the magnetic and kinetic control of advanced tokamak (AT) scenarios can be advantageously identified from simulated rather than real data, and used for control design. The method is applied to the combined control of the safety factor profile, q(x), and normalized pressure parameter, βN, using DIII-D parameters and actuators (on-axis co-current neutral beam injection (NBI) power, off-axis co-current NBI power, electron cyclotron current drive power, and ohmic coil). The approximate plasma response model was identified from simulated open-loop data obtained using a rapidly converging plasma transport code, METIS, which includes an MHD equilibrium and current diffusion solver, and combines plasma transport nonlinearity with 0D scaling laws and 1.5D ordinary differential equations. The paper discusses the results of closed-loop METIS simulations, using the near-optimal ARTAEMIS control algorithm (Moreau D et al 2013 Nucl. Fusion 53 063020) for steady state AT operation. With feedforward plus feedback control, the steady state target q-profile and βN are satisfactorily tracked with a time scale of about 10 s, despite large disturbances applied to the feedforward powers and plasma parameters. The robustness of the control algorithm with respect to disturbances of the H&CD actuators and of plasma parameters such as the H-factor, plasma density and effective charge, is also shown.

Integrated modeling of high βN steady state scenario on DIII-D

Theory-based integrated modeling validated against DIII-D experiments predicts that fully non-inductive DIII-D operation with βNu2009>u20094.5 is possible with certain upgrades. IPS-FASTRAN is a new iterative numerical procedure that integrates models of core transport, edge pedestal, equilibrium, stability, heating, and current drive self-consistently to find steady-state (d/dtu2009=u20090) solutions and reproduces most features of DIII-D high βN discharges with a stationary current profile. Projecting forward to scenarios possible on DIII-D with future upgrades, the high qminu2009>u20092 scenario achieves stable operation at βN as high as 5 by using a very broad current density profile to improve the ideal-wall stabilization of low-n instabilities along with confinement enhancement from low magnetic shear. This modeling guides the necessary upgrades of the heating and current drive system to realize reactor-relevant high βN steady-state scenarios on DIII-D by simultaneous optimization of the current and pressure profiles.