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In the initial stages of ITER operation, ELM mitigation systems need to be commissioned. This requires controlled flat-top operation in type-I ELMy H-mode regimes. Hydrogen or helium plasma discharges are used exclusively in these stages to ensure negligible production of neutrons from fusion reactions. With the expected higher L–H power threshold of hydrogen and helium plasmas compared to corresponding D and D/T plasmas, it is uncertain whether available auxiliary power systems are sufficient to operate in stable type-I ELMy H-mode. This has been investigated using integrated core and edge/SOL/divertor modelling with JINTRAC. Assuming that the L–H power threshold is well captured by the Martin08 scaling law, the presented simulations have found that 30 MW of ECRH power is likely required for the investigated hydrogen plasma scenarios, rather than the originally planned 20 MW in the 2016 Staged Approach ITER Baseline. However, past experiments have shown that a small helium fraction (∼10%) can considerably reduce the hydrogen plasma L–H power threshold. Assuming that these results extrapolate to ITER operation regimes, the 7.5 MA/2.65 T hydrogen plasma scenario is likely to access stable type-I ELMy H-mode operation also at 20 MW of ECRH.

JET experiments using the fuel mixture envisaged for fusion power plants, deuterium and tritium (D–T), provide a unique opportunity to validate existing D–T fusion power prediction capabilities in support of future device design and operation preparation. The 2021 JET D–T experimental campaign has achieved D–T fusion powers sustained over 5 s in ITER-relevant conditions i.e. operation with the baseline or hybrid scenario in the full metallic wall. In preparation of the 2021 JET D–T experimental campaign, extensive D–T predictive modelling was carried out with several assumptions based on D discharges. To improve the validity of ITER D–T predictive modelling in the future, it is important to use the input data measured from 2021 JET D–T discharges in the present core predictive modelling, and to specify the accuracy of the D–T fusion power prediction in comparison with the experiments. This paper reports on the validation of the core integrated modelling with TRANSP, JINTRAC, and ETS coupled with a quasilinear turbulent transport model (Trapped Gyro Landau Fluid or QualLiKiz) against the measured data in 2021 JET D–T discharges. Detailed simulation settings and the heating and transport models used are described. The D–T fusion power calculated with the interpretive TRANSP runs for 38 D–T discharges (12 baseline and 26 hybrid discharges) reproduced the measured values within 20 %. This indicates the additional uncertainties, that could result from the measurement error bars in kinetic profiles, impurity contents and neutron rates, and also from the beam-thermal fusion reaction modelling, are less than 20 % in total. The good statistical agreement confirms that we have the capability to accurately calculate the D–T fusion power if correct kinetic profiles are predicted, and indicates that any larger deviation of the D–T fusion power prediction from the measured fusion power could be attributed to the deviation of the predicted kinetic profiles from the measured kinetic profiles in these plasma scenarios. Without any posterior adjustment of the simulation settings, the ratio of predicted D–T fusion power to the measured fusion power was found as 65%–96% for the D–T baseline and 81%–97% for D–T hybrid discharge. Possible reasons for the lower D–T prediction are discussed and future works to improve the fusion power prediction capability are suggested. The D–T predictive modelling results have also been compared to the predictive modelling of the counterpart D discharges, where the key engineering parameters are similar. Features in the predicted kinetic profiles of D–T discharges such as underprediction of ne are also found in the prediction results of the counterpart D discharges, and it leads to similar levels of the normalized neutron rate prediction between the modelling results of D–T and the counterpart D discharges. This implies that the credibility of D–T fusion power prediction could be a priori estimated by the prediction quality of the preparatory D discharges, which will be attempted before actual D–T experiments. [ABSTRACT FROM AUTHOR]

STEP is a spherical tokamak prototype power plant that is being designed to demonstrate net electric power. The design phase involves the exploitation of plasma models to optimise fusion performance subject to satisfying various physics and engineering constraints. A modelling workflow, including integrated core plasma modelling, MHD stability analysis, SOL and pedestal modelling, coil set and free boundary equilibrium solvers, and whole plant design, has been developed to specify the design parameters and to develop viable scenarios. The integrated core plasma model JETTO is used to develop individual flat-top operating points that satisfy imposed criteria for fusion power performance within operational constraints. Key plasma parameters such as normalised beta, Greenwald density fraction, auxiliary power and radiated power have been scanned to scope the operational space and to derive a collection of candidate non-inductive flat-top points. The assumed auxiliary heating and current drive is either from electron cyclotron (EC) systems only or a combination of EC and electron Bernstein waves. At present stages of transport modelling, there is a large uncertainty in overall confinement for relevant parameter regimes. For each of the two auxiliary heating and current drive systems scenarios, two candidate flat-top points have been developed based on different confinement assumptions, totalling to four operating points. A lower confinement assumption generally suggests operating points in high-density, high auxiliary power regimes, whereas higher confinement would allow access to a broader parameter regime in density and power while maintaining target fusion power performance.

Designing a robust termination scenario for a burning ITER plasma is a challenge that requires extensive core plasma and divertor modelling. The presented work consists of coupled core/edge/SOL/divertor simulations, performed with the JINTRAC code, to study the Q = 10 flat-top phase and exit phase of the ITER 15 MA/5.3 T DT scenario. The modelling utilizes the recently implemented option to treat deuterium and tritium separately in the SOL/divertor, enabling a consistent treatment of deuterium and tritium in the whole plasma volume, which is a unique capability of JINTRAC. In addition, these are the first JINTRAC simulations of this scenario that use a first-principles transport model to self-consistently model the ECRH power deposition and to include tungsten while keeping track of tungsten sputtering and accumulation. The flat-top simulations demonstrate the possibility of sustaining a steady state fusion Q of 10 using pure deuterium gas puffs together with DT mixed pellets, which is an option to make a more effective use of tritium. Simulations of the exit phase are set up sequentially, with each phase providing initial conditions for the next, starting with a density decay at full current and auxiliary power, and demonstrate the possibility of reducing the density robustly within a few seconds. Following the density decay, a subsequent auxiliary power ramp-down in H-mode is performed with a late H–L transition at low auxiliary power, which may provide an option for the optimization of the plasma termination. The final ramp-down phase consists of a current ramp-down in L-mode to 3.75 MA.

Experimentally, it has been observed in high-confinement (H-Mode) plasmas with Edge Localised Modes (ELMs) on JET that the pressure pedestal (pe,ped) is degraded by approximately a factor of two when there is a change in electron separatrix density, ne,sep, from 1−4×1019m−3. Previous work using the pedestal stability code EUROPED, has been able to predict the degradation of pe,ped but only for ne,sep≤1.5×1019m−3. In this work, we apply a coupled code JETTO-MISHKA-FRANTIC, to self-consistently predict the transport in the pedestal region and neutral source with varying separatrix conditions. The code feeds back on the transport in the pedestal region to achieve profiles that are marginally stable to ideal MHD modes (continuous ELM model in JETTO).When accounting for the change in electron separatrix temperature (Te,sep), ion separatrix temperature (Ti,sep) and the poloidally integrated neutral flux crossing the separatrix (Γsep,neut) as it changes with ne,sep (according to a scan in ne,sep in the edge code EDGE2D-EIRENE), no degradation in pe,ped was observed in JETTO-MISHKA-FRANTIC in contrast to experiment. Instead, an increase in pe,ped with ne,sep was observed which is driven by an increasing density pedestal (ne,ped). Within the presented JETTO-MISHKA-FRANTIC simulations, changing the pedestal width by a factor of two and a half in normalised poloidal flux (ψn) resulted in an approximately 40% degradation in pe,ped for ne,sep=1−3×1019m−3. This change in pedestal width was not supported by experimental data. A scan in the ratio of particle and energy transport in the pedestal (D/χ) was found to have a negligible effect on pe,ped. Qualitative agreement between JETTO-MISHKA-FRANTIC with EUROPED was found when the input density profiles are identical.

Simulations of JET ITER-like wall high-confinement mode plasmas, including type-I edge-localised modes (ELMs), using JINTRAC for the background plasmas and ERO2.0 for tungsten erosion and transport, predict virtually perfect screening of the primary W erosion sources at the divertor targets during both the ELM and inter-ELM phases. The largest source of W influx to the main plasma is predicted to be the outer vertical divertor due to sputtering by energetic fuel (D, T) atoms from charge-exchange reactions. ERO2.0 predictions accurately reproduce the measured W I emission in the low-field side divertor, but underpredict the W II emission by a factor of 10. Potential reasons for the W II discrepancy include uncertainties in the atomic data, assumptions on the sheath properties and the sputtering angle distribution, and the impact of metastable states.

Pellet injection is regarded as the only realistic actuator for core density control in future reactors such as ITER and DEMO. However, a control strategy that can reliably regulate the plasma close to operational limits using multiple pellet injectors is not yet available. In this paper, we present the first integrated model control simulations where a dedicated model-predictive controller is included in JINTRAC. We show that, when continuous actuators are considered, a simple transport model with a steady-state disturbance rejection paradigm is capable of capturing the particle transport dynamics for multiple transport models and scenarios. This in turn allows the model-predictive controller to deal with the uncertainty and minimize the control error given the limited actuation space. Furthermore, we show that for ITER and DEMO relevant pellet sizes, the discrete, nonlinear dynamics of pellet injection will limit the control performance and jeopardize the constraints if not accounted for by the controller. Hence, we conclude that for high-performance control on future reactors, controllers will have to be developed that explicitly deal with the discrete pellet dynamics.