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.

Abstract The advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at an aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2, and triangularity of 0.63. The broadest pressure cases reached wall-stabilized βN ∼ 5.75, limited by n = 3 external kink mode requiring a conducting shell at b/a = 0.3, requiring plasma rotation, feedback, and/or kinetic stabilization. The medium pressure peaking case reaches βN = 5.28 with BT = 6.75, while the peaked pressure case reaches βN < 5.15. Fast particle magnetohydrodynamic stability shows that the alpha particles are unstable, but this leads to redistribution to larger minor radius rather than loss from the plasma. Edge and divertor plasma modeling shows that 75% of the power to the divertor can be radiated with an ITER-like divertor geometry, while =95% can be radiated in a stable detached mode with an orthogonal target and wide slot geometry. The bootstrap current fraction is 91% with a q95 of 4.5, requiring ∼1.1 MA of external current drive. This current is supplied with 5 MW of ion cyclotron radio frequency/fast wave and 40 MW of lower hybrid current drive. Electron cyclotron is most effective for safety factor control over ρ ∼ 0.2 to 0.6 with 20 MW. The pedestal density is ∼0.9×1020/m3, and the temperature is ∼4.4 keV. The H98 factor is 1.65, n/nGr = 1.0, and the ratio of net power to threshold power is 2.8 to 3.0 in the flattop.

Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a βN of 5.75, H98 of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βN of 2.5, H98 of 1.25, n/nGr of 1.3, and peak divertor heat flux of 10 MW/m. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.

Physics based integrated modelling of the baseline scenario for a Fusion Nuclear Science Facility based on the Advanced Tokamak concept (FNSF-AT) (Chan et al 2010 Fusion Sci. Technol. 57 66) has found steady-state equilibria with good stability and controllability properties at the fusion performance required to accomplish FNSF's nuclear science mission with margin. 2D divertor analysis for this baseline scenario predicts that peak heat flux <10 MW m(-2) can be obtained even with scrape-off layer power width similar to 1 mm. Using this baseline fusion performance, high fidelity and high-resolution 3D neutronics calculations show acceptable cumulative end-of-life organic insulator dose levels in all the device coils, and TBR > 1. Two current drive scenarios, two divertor configurations, and two blanket concepts have been analysed. FNSF-AT would complement ITER in addressing science and technology gaps to a commercially attractive DEMO, and could enable a DEMO construction decision triggered by the achievement of Q = 10 in ITER.

Recent progress on ITER steady-state (SS) scenario modelling by the ITPA-IOS group is reviewed. Code-to-code benchmarks as the IOS group's common activities for the two SS scenarios (weak shear scenario and internal transport barrier scenario) are discussed in terms of transport, kinetic profiles, and heating and current drive (CD) sources using various transport codes. Weak magnetic shear scenarios integrate the plasma core and edge by combining a theory-based transport model (GLF23) with scaled experimental boundary profiles. The edge profiles (at normalized radius ρ = 0.8–1.0) are adopted from an edge-localized mode-averaged analysis of a DIII-D ITER demonstration discharge. A fully noninductive SS scenario is achieved with fusion gain Q = 4.3, noninductive fraction f NI = 100%, bootstrap current fraction f BS = 63% and normalized beta βN = 2.7 at plasma current I p = 8 MA and toroidal field B T = 5.3 T using ITER day-1 heating and CD capability. Substantial uncertainties come from outside the radius of setting the boundary conditions (ρ = 0.8). The present simulation assumed that βN (ρ) at the top of the pedestal (ρ = 0.91) is about 25% above the peeling–ballooning threshold. ITER will have a challenge to achieve the boundary, considering different operating conditions (T e/T i ≈ 1 and density peaking). Overall, the experimentally scaled edge is an optimistic side of the prediction. A number of SS scenarios with different heating and CD mixes in a wide range of conditions were explored by exploiting the weak-shear steady-state solution procedure with the GLF23 transport model and the scaled experimental edge. The results are also presented in the operation space for DT neutron power versus stationary burn pulse duration with assumed poloidal flux availability at the beginning of stationary burn, indicating that the long pulse operation goal (3000 s) at I p = 9 MA is possible. Source calculations in these simulations have been revised for electron cyclotron current drive including parallel momentum conservation effects and for neutral beam current drive with finite orbit and magnetic pitch effects.

The objective of the Fusion Development Facility (FDF) under consideration is to carry forward advanced tokamak physics for optimization of fusion reactors and enable development of fusion's energy applications. A concept of FDF based on the tokamak approach with conservative expressions of advanced physics and nonsuper-conducting magnet technology is presented. It is envisioned to nominally provide 2 MW/m(2) of neutron wall loading and operate continuously for up to 2 weeks as required for fusion nuclear component research and development. FDF will have tritium breeding capability with a goal of addressing the tritium self-sufficiency issue for fusion energy. A zero-dimensional system study using extrapolations of current physics and technology is used to optimize FDF for reasonable power consumption and moderate size. It projects a device that is between the DIII-D tokamak (major radius 1.8 m) [J. L. Luxon, Nucl. Fusion, Vol. 42, p. 614 (2002)] and the Joint European Torus (major radius 3 m) [P. H. Rebut, R. J. Bickerton, and B. E. Keen, Nucl. Fusion, Vol. 25, p. 1011 (1985)] in size, with an aspect ratio A of 3.5 and a fusion gain Q of 2 to 5. Theory-based stability and transport modeling is used to complement the system study and to address physics issues related to specific design points. It is demonstrated that the FDF magnetohydrodynamic stability limits can be readily met with conservative stabilizing conducting wall placement. Transport analysis using a drift-wave-based model with an edge boundary condition consistent with the pedestal stability limit indicates that the FDF confinement requirement can also be readily satisfied. A surprising finding is that the toroidal Alfven eigenmodes are stabilized by strong ion Landau damping. Analysis of vertical stability control indicates that the basis configuration with an elongation kappa(X) similar to 2.35 can be controlled using a power supply technology similar to that used in DIII-D. Peak heat fluxes to the divertor are somewhat lower than those of ITER (R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Control. Fusion, Vol. 44, p. 519 (2002)), but FDF will operate with a higher duty factor.

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, B T, and the plasma current, I p, 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 B T 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.

Neutral beam current drive (NBCD) experiments in DIII-D using vertically shifted plasmas to move the current drive away from the axis have clearly demonstrated robust off-axis NBCD. Time-dependent measurements of magnetic field pitch angles by the motional Stark effect diagnostic are used to obtain the evolution of the poloidal magnetic flux, which indicates a broad off-axis NBCD profile with a peak at about half the plasma minor radius. In most cases, the measured off-axis NBCD profile is consistent with calculations using an orbit-following Monte Carlo code for the beam ion slowing down including finite-orbit effects provided there is no large-scale magnetohydrodynamic activity such as Alfvén eigenmodes modes or sawteeth. An alternative analysis method shows good agreement between the measured pitch angles and those from simulations using transport-equilibrium codes. Two-dimensional image of Doppler-shifted fast ion Dα light emitted by neutralized energetic ions shows clear evidence for a hollow profile of beam ion density, consistent with classical beam ion slowing down. The magnitude of off-axis NBCD is sensitive to the alignment of the beam injection relative to the helical pitch of the magnetic field lines. If the signs of toroidal magnetic field and plasma current yield the proper helicity, both measurement and calculation indicate that the efficiency is as good as on-axis NBCD because the increased fraction of trapped electrons reduces the electron shielding of the injected ion current, in contrast with electron current drive schemes where the trapping of electrons degrades the efficiency. The measured off-axis NBCD increases approximately linearly with the injection power, although a modest amount of fast ion diffusion is needed to explain an observed difference in the NBCD profile between the measurement and the calculation at high injection power.