A new theory-based transport model with comprehensive physics (trapping, general toroidal geometry, fully electromagnetic, electron-ion collisions, impurity ions) has been developed. The core of the model is the new trapped-gyro-Landau-fluid (TGLF) equations, which provide a fast and accurate approximation to the linear eigenmodes for gyrokinetic drift-wave instabilities (trapped ion and electron modes, ion and electron temperature gradient modes, and kinetic ballooning modes). The new TGLF transport model is more accurate, and has an extended range of validity, compared to its predecessor GLF23. The TGLF model unifies trapped and passing particles in a single set of gyro-Landau-fluid equations. A model for the averaging of the Landau resonance by the trapped particles makes the equations work seamlessly over the whole drift-wave wave-number range from trapped ion modes to electron temperature gradient modes. A fast eigenmode solution method enables unrestricted magnetic geometry. The transport model uses...

http://absimage.aps.org/image/DPP06/MWS_DPP06-2006-000826.pdf

A Theory-Based Transport Model With Comprehensive Physics

Since the last IAEA-FEC in 2016, the EAST physics experiments have been developed further in support of high-performance steady-state operation for ITER and CFETR. First demonstration of g100 seconds time scale long-pulse steady-state scenario with a good plasma performance (H98(y2) ~ 1.1) and a good control of impurity and heat exhaust with the upper tungsten divertor has been achieved on EAST using the pure radio frequency (RF) power heating and current drive. The EAST operational domain has been significantly extended towards more ITER and CFETR related high beta steady-state regime (βP ~ 2.5 & βN ~ 1.9 of using RF & NB and βP ~ 1.9 & βN ~ 1.5 of using pure RF). A large bootstrap current fraction up to 47% has been achieved with with q95~6.0-7.0. The interaction effect between the electron cyclotron resonant heating (ECRH) and two lower hybrid wave (LHW) systems has been investigated systematically, and applied for the improvement of current drive efficiency and plasma confinement quality in the steady-state scenario development on EAST. Full ELM suppression using the n= 2 RMPs has been achieved in ITER-like standard type-I ELMy H-mode plasmas with a range of the edge safety factor of q95 ≈ 3.2-3.7 on EAST. Reduction of the peak heat flux on the divertor was demonstrated using the active radiation feedback control. An increase in the total heating power and improvement of the plasma confinement are expected using a 0-D model prediction for higher bootstrap fraction. Towards long-pulse, high bootstrap current fraction operation, a new lower ITER-like tungsten divertor with active water-cooling will be installed, together with further increase and improvement of heating and current drive capability. &#13; &#13;

/pdf/recent-advances-in-east-physics-experiments-in-support-of-3ifpzd7apd.pdf

Recent advances in EAST physics experiments in support of steady-state operation for ITER and CFETR

An energy principle is used to obtain the solution of the magnetohydrodynamic (MHD) equilibrium equation J×B−∇p=0 for nested magnetic flux surfaces that are expressed in the inverse coordinate representation x=x(ρ, θ, ζ). Here, θ are ζ are poloidal and toroidal flux coordinate angles, respectively, and p=p(ρ) labels a magnetic surface. Ordinary differential equations in ρ are obtained for the Fourier amplitudes (moments) in the doubly periodic spectral decomposition of x. A steepest‐descent iteration is developed for efficiently solving these nonlinear, coupled moment equations. The existence of a positive‐definite energy functional guarantees the monotonic convergence of this iteration toward an equilibrium solution (in the absence of magnetic island formation). A renormalization parameter λ is introduced to ensure the rapid convergence of the Fourier series for x, while simultaneously satisfying the MHD requirement that magnetic field lines are straight in flux coordinates. A descent iteration is also developed for determining the self‐consistent value for λ.

Steepest descent moment method for three-dimensional magnetohydrodynamic equilibria

Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance {approx}10 has been sustained for >5 {tau}{sub E} with q{sub min} >1.5. (The normalized performance is measured by the product {beta}{sub N} H{sub 89} indicating the proximity to the conventional {beta} limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and {beta} {approx}{le} 5%. The limit to increasing {beta} is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate {approx}75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and {beta} control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with {beta}{sub N}H{sub 89} {approx} 7 for up to 6.3 s or {approx} 34 {tau}{sub E}. These discharges appear to be in resistive equilibrium with q{sub min} {approx} 1.05, in agreement with the current profile relaxation time of 1.8 s.

/pdf/long-pulse-high-performance-discharges-in-the-diii-d-tokamak-6dams62vbg.pdf

Long-pulse, high-performance discharges in the diii-d tokamak

With the first tokamak designed for full nuclear operation now well into final assembly (ITER), and a major new research tokamak starting commissioning (JT60SA), nuclear fusion is becoming a mainstream potential energy source for the future. A critical part of the viability of magnetic confinement for fusion is superconductor technology. The experience gained and lessons learned in the application of this technology to ITER and JT60SA, together with new and improved superconducting materials, is opening multiple routes to commercial fusion reactors. The objective of this roadmap is, through a series of short articles, to outline some of these routes and the materials/technologies that go with them.

Superconductors for fusion: A roadmap

The Chinese Fusion Engineering Testing Reactor (CFETR), complementing the ITER facility, is aiming to demonstrate fusion energy production up to 200 MW initially and to eventually reach DEMO relevant power level 1 GW, to manifest a high duty factor of 0.3–0.5, and to pursue tritium self-sufficiency with tritium breeding ratio (TBR) >1. The key challenge to meet the missions of the CFETR is to run the machine in steady state (or long pulse) and high duty factor. By using a multi-dimensional code suite with physics-based models, self-consistent steady-state and hybrid mode scenarios for CFETR have been developed under a high magnetic field up to 6.5 T. The negative-ion neutral beam injection together with high frequency electron cyclotron wave and lower hybrid wave (and/or fast wave) are proposed to be used to drive the current. Subsequently the engineering design of CFETR including the magnet system, vacuum system, tritium breeding blanket, divertor, remote handling and maintenance system will be introduced. Some research and development (R&D) activities are also introduced in this paper.

https://iopscience.iop.org/article/10.1088/1741-4326/ab0e27/pdf

Progress of the CFETR design

https://iopscience.iop.org/article/10.1088/1741-4326/abaf33/pdf

Progress in extending high poloidal beta scenarios on DIII-D towards a steady-state fusion reactor and impact of energetic particles

https://iopscience.iop.org/article/10.1088/1741-4326/abaa91/pdf

Experimental progress of hybrid operational scenario on EAST tokamak

http://absimage.aps.org/image/DPP14/MWS_DPP14-2014-000472.pdf

Faraday-Effect Polarimeter-Interferometer System for current density measurement on EAST

https://iopscience.iop.org/article/10.1088/1741-4326/ac1546/pdf

G.Q. Li

Papers

Progress of the CFETR design

Progress in extending high poloidal beta scenarios on DIII-D towards a steady-state fusion reactor and impact of energetic particles

Experimental progress of hybrid operational scenario on EAST tokamak

Faraday-Effect Polarimeter-Interferometer System for current density measurement on EAST

Effects of radial transport on divertor power and particle flux widths under different operational regimes in EAST