In this work we present the extensive validation of a refined version of the integrated model based on engineering parameters (IMEP) introduced in reference (Ludaet al2020Nucl .Fusion60036023). The modeling workflow is now fully automated, computationally faster thanks to the reduced radial resolution of the TGLF calculation, and it includes the modeling of the toroidal rotation, which was still taken from experimental measurements in our previous work. The updated model maintains the same accuracy as its previous version when tested on the cases presented in the initial publication. The confined plasma, from the magnetic axis to the separatrix, is simulated without using any experimental information from profiles measurements, and the inputs of IMEP are the same engineering parameters used when programming a plasma discharge. The model validation database consists of 50 ASDEX Upgrade (AUG) stationary (over a few energy confinement time) H-mode phases, which largely cover the entire AUG operational domain. The prediction of IMEP is compared with experimental measurements and with scaling laws, such as the IPB98(y,2), the ITPA20-IL, and AUG specific regressions. This modeling framework has proven to be very accurate over the entire set of 50 cases, with a significantly lower mean relative error with respect to each of the scaling laws considered, accurately reproducing the change in pedestal and core confinement caused by a change in plasma current, heating power, fueling rate, triangularity, magnetic field, NBI voltage (i.e. the effect of a change in the core particle source), and heating mix (e.g. correctly predicting the effect on confinement caused by a change inT _e/ T _i). Plasma confinement is correctly described by IMEP also for two particular operating regimes, such as the ITER baseline scenario, and the QCE regime (quasi continuous exhaust, also referred as type-II and small ELMs). This work clearly demonstrates the power of this approach in pulling out physics mechanisms to interpret subtle interdependencies and that a 1D integrated model can reproduce experimental results over very large parameter variations with a higher accuracy than any statistical regression. This approach has therefore the potential to improve the prediction of the fusion performance in future tokamak reactors.

A new lower tungsten divertor has been developed and installed in the EAST superconducting tokamak to replace the previous graphite divertor with power handling capability increasing from <2 MW m⁻² to ∼10 MW m⁻², aiming at achieving long-pulse H-mode operations in a full metal wall environment with the steady-state divertor heat flux of ∼10 MW m⁻². A new divertor concept, 'corner slot' (CS) divertor, has been employed. By using the 'corner effect', a strongly dissipative divertor with the local buildup of high neutral pressure near the corner can be achieved, so that stable detachment can be maintained across the entire outer target plate with a relatively lower impurity seeding rate, at a separatrix density compatible with advanced steady-state core scenarios. These are essential for achieving efficient current drive with low-hybrid waves, a low core impurity concentration and thus a low loop voltage for fully non-inductive long-pulse operations. Compared with the highly closed small-angle-slot divertor in DIII-D, the new divertor in EAST exhibits the following merits: (1) a much simpler geometry with integral cassette body structure, combining vertical and horizontal target plates, which are more suitable for actively water-cooled W/Cu plasma facing components, facilitating installation precision control for minimizing surface misalignment, achieving high engineering reliability and lowering the capital cost as well; (2) it has much greater flexibility in magnetic configurations, allowing for the position of the outer strike point on either vertical or horizontal target plates to accommodate a relatively wide triangularity range,δ _l = 0.4–0.6, thus enabling to explore various advanced scenarios. A water-cooled copper in-vessel coil has been installed under the dome. Five supersonic molecular beam injection systems have been mounted in the divertor to achieve faster and more precise feedback control of the gas injection rate. Furthermore, this new divertor allows for double null divertor operation and slowly sweeping the outer strike point across the horizontal and vertical target plates to spread the heat flux for long-pulse operations. Preliminary experimental results demonstrate the 'corner effect' and are in good agreement with simulations using SOLPS-ITER code including drifts. The EAST new divertor provides a test-bed for the closed divertor concept to achieve steady-state detachment operation at high power. Next step, a more closed divertor, 'sharp-cornered slot' divertor, building upon the current CS divertor concept, has been proposed as a candidate for the EAST upper divertor upgrade.

The efficient operation of a tokamak is limited by several constraints, such as the transition to high confinement or the density limits occurring in both confinement regimes. These particular boundaries of operation are derived in terms of a combination of dimensionless parameters describing interchange-drift-Alfvén turbulence without any free adjustable parameter. The derived boundaries describe the operational space at the separatrix of the ASDEX Upgrade tokamak, which is presented in terms of an electron density and temperature existence diagram. The derived density limits are compared against Greenwald scaling. The power threshold and role of ion heat flux for the transition to high confinement are discussed.

Magnetic reconnection in W7-X is studied by means of global numerical simulations in a series of models of increasing complexity. The magnetic geometry ranges from that of a cylinder to the full three-dimensional field of W7-X, and the equations solved range from ideal magnetohydrodynamics (MHD) to gyrokinetics. We simulate plasmas from the first operation phase with electron cyclotron current drive (ECCD). These are characterized by an equilibrium magnetic field featuring an ECCD-distorted 'humped' profile of the rotational transformι , withι= 1 in two radial locations. Such plasmas generally show sawtooth activity, hence motivating the present study. We pay particular attention to the role of equilibrium current density gradients in the destabilization of reconnecting modes. When the equilibrium temperature and density gradients are artificially suppressed (to eliminate the pressure gradient drive), the perturbed electrostatic potential is radially localized between the locations at whichι= 1. This is shown with a purely collisionless gyrokinetic model, in cylindrical geometry. In the real toroidal geometry of W7-X, for a non-ideal MHD model including a uniform resistivity, electron inertia and (numerical) viscosity, the same qualitative behaviour is observed. In particular, even if a resonant ( m ,n ) = (1, −1) perturbation is initialized, the most unstable mode is the ( m ,n ) = (−4, 4), wheremandnare the poloidal and toroidal mode numbers, respectively. Other modes are destabilized due to geometric coupling. The growth rate of this instability scales asη ^1/3, whereηis the plasma resistivity, thus suggesting that ECCD drives ideal MHD stable W7-X plasmas towards non-ideal marginality. An ideal magnetohydrodynamic analysis confirms the result. A fluid-kinetic hybrid version of theEUTERPEcode shows that gyrokinetic ions have a stabilizing effect on these modes. For W7-X relevant collisionalities, the growth rate scales linearly with the electron skin depth,d _e. Implications of our results for sawtoothing W7-X operation are discussed.

Here, we report on role of pre-ionization using non-resonant 2.45 GHz microwave heating ( P _{2.45 GHz} < 20 kW) in plasma start-up of neutral beam injection (NBI) for heliotron configurations in low beam power ( P _NB) under non-resonant heating condition. A rapid electron heating towards burn-through of the low- Zimpurities was observed experimentally in the early phase of beam injection when the seed plasma density produced by the non-resonant heating was enough for the plasma start-up. Beam heating time to the burn-through increased with decreasing the seed plasma density and a critical density condition of the seed plasma for successful start-up was observed experimentally. Proper timing of the gas fuelling is critical for plasma expansion because the beam fuelling is not significant. A 0-dimensional (0D) model analysis of the NBI start-up developed in this study well reproduces the experimental results. The 0D model clarifies the physical mechanism of the NBI start-up using pre-ionization described as follows: (1) the seed plasma produces sufficient beam ions immediately after beam injection, (2) the beam ions heat up electrons that promote the ionization/dissociation of the background neutrals, (3) this process acts as a positive feedback loop resulting in further electron heating towards burn-through. The 0D model analysis shows that the critical density corresponds to the state at which the electron heating by the beam ions is equal to electron power loss due to conduction and ionization/dissociation.

As a complement to our recent work [ 1 ], which focused on understanding the basic detachment physics and general experimental and numerical trends observed in the W7-X island divertor, this paper compares EMC3-Eirene simulation results with different local diagnostics, including IR cameras, Langmuir probes, H_α-cameras, and Thomson scattering. The main purposes are to (1) justify the simulation setup in the previous work, (2) identify the application limitations of the current EMC3-Eirene model, (3) verify the consistency of different diagnostics, and (4) isolate the main geometric and physical effects that need to be prioritized in further developing the EMC3-Eirene code and improving diagnostic capabilities. It turns out that the current version of the EMC3-Eirene code (without drifts) is not yet able to quantitatively reproduce all selected local measurements simultaneously under the current experimental conditions (in particular, the existence of error fields). Nevertheless, it can be shown that within a reasonable range of variation in magnetic configuration, cross-field transport, and SOL plasma state in the modeling, a region of overlap between the numerical results and the local measurements can be established. More accurate model-experiment comparisons will require clarification of error fields, implementation of drifts in EMC3, and improvement of diagnostic capabilities.