Long Pulse Scenarios (LPS) in ITER foreseen during the Pre-Fusion Power Operation (PFPO) phase of the ITER Research Plan (IRP) are assessed using 1.5D transport simulations within the ASTRA framework. Such assessment is required to predict the operational space for LPS operation in PFPO, as well as to evaluate which physics processes for LPS operation during Fusion Power Operation (FPO) could be studied during PFPO. An important aspect in the development of LPSs in PFPO is to minimize lifetime consumption of the Central Solenoid (CS) for these scenarios. The maximum pulse length achievable for LPSs in PFPO with no consumption of CS lifetime (currents in CS coils ⩽30 kA per turn) has been assessed for a range of heating schemes and heating mixes, confinement regimes (L-mode and H-mode) and for helium and hydrogen plasmas. The operational space of LPS and pulse length has been explored through density scans with the Heating and Current Drive mix required for the FPOQ⩾ 5 steady-state plasma scenario (namely Neutral Beam Injection and Electron Cyclotron Heating) including acceptable shine through losses on the first wall for both helium and hydrogen plasmas. Fast particle physics aspects that are common between FPO plasmas and LPS PFPO H-mode plasmas at low densities are studied including MHD stability analysis with the KINX code and non-perturbative critical gradient model based on high-n Toroidal Alfven Eigenmodes (TAE) stability kinetic ballooning code HINST calculations.

Pellet injection is the most promising technique to achieve efficient plasma core fuelling, key for attaining stationary scenarios in large magnetic confinement fusion devices. In this paper, the injection of pellets with different volumes and speeds into standard plasma scenarios in ITER (tokamak) and Wendelstein 7-X (stellarator) is studied by modeling the pellet ablation and particle deposition, focusing on the evaluation of the expected differences in pellet plasmoid drifts in tokamaks and stellarators. Since the efficiency of the damping-drift mechanisms is predicted to depend on the magnetic configuration, device-specific characteristics are expected for the temporal evolution of the plasmoid drift acceleration. For instance, plasmoid-internal Pfirsch–Schlüter currents dominate the drift damping process for stellarators, while plasmoid-external currents are more relevant for tokamaks. Also, relatively larger drifts are in principle expected for W7-X due to higher field gradients in relation to machine dimensions. However, shorter plasmoid-internal charge reconnection lengths result in the drift damping due to internal Pfirsch–Schlüter currents being more effective than in a tokamak. Therefore, the average relative drift displacement during the whole plasmoid homogenization maya prioribe comparable in both magnetic configurations. Moreover, High Field Side (HFS) injection is expected to be highly advantageous to maximize pellet particle deposition in ITER, whereas it may only be beneficial in medium to highβenvironments in W7-X. Finally, there may be means for the optimization of pellet injection configurations in both ITER and W7-X for the considered plasma scenarios despite the sizeable differences in the relative importance of the mechanisms of plasmoid drift acceleration and deceleration in play.

Theq ₉₅ window for Type-I edge localized modes (ELMs) suppression usingn  = 4 even parity resonant magnetic perturbations (RMPs) has been significantly expanded to the ranges [3.9, 4.1] and [4.2, 4.8] in EAST while maintaining good confinement, which is demonstrated to be reliable and repeatable over the last two years. This window is significantly wider than the previous one achieved usingn  = 4 odd parity RMPs, which is around q_{95} = 3.7pm0.1 . Here,nrepresents the toroidal mode number of the applied RMPs andq ₉₅ is the safety factor at 95%of the normalized poloidal magnetic flux. During ELM suppression, there is only a slight drop in the plasma stored energy and density ({leqslant}10% ). The comparison of changes in the pedestal profiles suggests that ELM suppression is achieved when the pedestal gradient is kept lower than a threshold. This wideq ₉₅ window for ELM suppression is consistent with the prediction made by MARS-F modeling prior to the experiment, which located it at one of the resonantq ₉₅ windows for plasma response. The Chirikov parameter taking into account plasma response near the pedestal top, which measures plasma edge stochasticity, significantly increases whenq ₉₅ exceeds 4, mainly due to the denser neighboring rational surfaces. The modeling of plasma response reveals a strong coupling between resonant and non-resonant components across the pedestal region, which is a characteristic of the kink-peeling like response observed during RMP-ELM suppression in previous studies on EAST. These promising results demonstrate the reliability of ELM suppression using then  = 4 RMPs in EAST and expand the physical understanding on ELM suppression mechanism.

Screening of high-Z (W) impurities from the confined plasma by the temperature gradient at the plasma periphery of fusion-grade H-mode plasmas has been demonstrated in the JET-ILW (ITER-like wall) tokamak. Through careful optimisation of the hybrid-scenario, deuterium plasmas with sufficient heating power (gtrsim32 MW), high enough ion temperature gradients at the H-mode pedestal top can be achieved for the collisional, neo-classical convection of the W impurities to be directed outwards, expelling them from the confined plasma. Measurements of the W impurity fluxes between and during edge-localised modes (ELMs) based on fast bolometry measurements show that in such plasmas there is a net efflux (loss) between ELMs but that ELMs often allow some W back into the confined plasma. Provided steady, high-power heating is maintained, this mechanism allows such plasmas to sustain high performance, with an average D–D neutron rate of {sim} 3.2 imes 10^{16}  s⁻¹ over a period of ∼3 s, after an initial overshoot (equivalent to a D–T fusion power of ∼9.4 MW), without an uncontrolled rise in W impurity radiation, giving added confidence that impurity screening by the pedestal may also occur in ITER, as has previously been predicted (Duxet al2017Nucl. Mater. Energy1228–35).

Neoclassical toroidal viscosity (NTV) torque caused by resonant magnetic perturbation (RMP) and the induced toroidal momentum transport are investigated for International Thermonuclear Experimental Reactor (ITER) scenarios through numerical modeling. The NTV torque is calculated using the NTVTOK code including the bounce-drift resonant effect, and the toroidal rotation evolution is modeled by solving a toroidal momentum transport equation that couples momentum source (NTV torque) and the momentum diffusion effect. The variation of RMP coil phasing (defined as toroidal phase difference between different rows of RMP coils) results in different types of plasma response and hence different features of NTV torque and toroidal momentum transport. The bounce-drift resonant effect enhances NTV torque and induces more significant toroidal rotation variation than simulations that adopt the bounce-averaged NTV model. With the initial rotation of the ITER design, plasma rotation is braked by NTV torque, but it may be sustained at moderate amplitude due to electron contributions to NTV torque. It is also found that initially static or slowly rotating plasma can be accelerated by NTV torque either toward co-I_mathrm por counter-I_mathrm p(I_mathrm pindicates plasma current) direction, indicating that NTV torque can be regarded as a momentum source for plasma with low torque injection; for instance, radio-frequency heated plasma.