Core ( r / a  lt  0.5) toroidal rotation from argon (Ar¹⁷⁺, 40 AMU) and molybdenum (Mo³²⁺, 96 AMU) ions has been compared in C-Mod tokamak plasmas over a wide range of operating conditions and confinement schemes, including Ohmic L-mode in the linear and saturated regimes, ion cyclotron range of frequencies heated I-mode and H-mode, as well as in discharges with induced locked modes and with external current and rotation drive. In all cases the velocities of the two impurities are identical within about 5%, for a range between −60 and +80 km s⁻¹. This is in general agreement with the predictions of neo-classical theory.

The dependence of the confinement of a tokamak plasma in L-mode on the magnetic field is explored with a set of dedicated experiments in ASDEX Upgrade and with a theory-based full-radius modelling approach, based on the ASTRA transport code and the TGLF-SAT2 transport model and only using engineering parameters in input, like those adopted in scaling laws for the confinement time. The experimental results confirm the weak dependence of the global confinement on the magnetic field, consistent with the scaling laws for L-mode plasmas and in agreement with the full-radius TGLF-SAT2 predictions. The modelling approach is then extended to numerically investigate the confinement dependence on magnetic field, plasma current and plasma size. The weak dependence of the L-mode confinement on the magnetic field at constant plasma current and plasma size is shown to be produced by a balance between the decrease of confinement mainly produced by the reduction of theE×Bshearing rate and the increase of confinement provided by the reduced gyro-Bohm factor, when the magnetic field is increased. The ASTRA/TGLF-SAT2 predicted increase of confinement with increasing plasma size is investigated in comparison with the predictions of the global confinement scaling laws for L-mode plasmas and the Bohm and gyro-Bohm dependencies of confinement, highlighting interesting similarities and important differences. Full-radius TGLF-SAT2 simulations with increasing plasma size are then extended to dimensions which are compatible with reactor relevant fusion power production, using ITER and the European DEMO as references. ASTRA/TGLF-SAT2 predictions of fusion power and confinement of an L-mode fusion reactor are presented at both 5.7 T and 10 T of magnetic field on the magnetic axis.

As part the DTE2 campaign in the JET tokamak, we conducted a parameter scan in T and D-T complementing existing pulses in H and D. For the different main ion masses, type-I ELMy H-modes at fixed plasma current and magnetic field can have the pedestal pressure varying by a factor of 4 and the total pressure changing from beta_{mathrm{N}} = 1.0to 3.0. We investigated the pedestal and core isotope mass dependencies using this extensive data set. The pedestal shows a strong mass dependence on the density, which influences the core due to the strong coupling between both plasma regions. To better understand the causes for the observed isotope mass dependence in the pedestal, we analysed the interplay between heat and particle transport and the edge localised mode (ELM) stability. For this purpose, we developed a dynamic ELM cycle model with basic transport assumptions and a realistic neutral penetration. The temporal evolution and resulting ELM frequency introduce an additional experimental constraint that conventional quasi-stationary transport analysis cannot provide. Our model shows that a mass dependence in the ELM stability or in the transport alone cannot explain the observations. One requires a mass dependence in the ELM stability as well as one in the particle sources. The core confinement time increases with pedestal pressure for all isotope masses due to profile stiffness and electromagnetic turbulence stabilisation. Interestingly, T and D-T plasmas show an improved core confinement time compared to H and D plasmas even for matched pedestal pressures. For T, this improvement is largely due to the unique pedestal composition of higher densities and lower temperatures than H and D. With a reduced gyroBohm factor at lower temperatures, more turbulent drive in the form of steeper gradients is required to transport the same amount of heat. This picture is supported by quasilinear flux-driven modelling usingTGLF -SAT2 withinAstra . With the experimental boundary conditionTGLF -SAT2 predicts the core profiles well for gyroBohm heat fluxes {gt}15 , however, overestimates the heat and particle transport closer to the turbulent threshold.

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).

The JET hybrid scenario has been developed from low plasma current carbon wall discharges to the record-breaking Deuterium-Tritium plasmas obtained in 2021 with the ITER-like Be/W wall. The development started in pure Deuterium with refinement of the plasma current, and toroidal magnetic field choices and succeeded in solving the heat load challenges arising from 37 MW of injected power in the ITER like wall environment, keeping the radiation in the edge and core controlled, avoiding MHD instabilities and reaching high neutron rates. The Deuterium hybrid plasmas have been re-run in Tritium and methods have been found to keep the radiation controlled but not at high fusion performance probably due to time constraints. For the first time this scenario has been run in Deuterium-Tritium (50:50). These plasmas were re-optimised to have a radiation-stable H-mode entry phase, good impurity control through edgeT_igradient screening and optimised performance with fusion power exceeding 10 MW for longer than three alpha particle slow down times, 8.3 MW averaged over 5 s and fusion energy of 45.8 MJ.