A new method for localizing the perturbations based on knowledge of their amplitude is briefly presented. Relying on the successful testing of the perturbations model (Miron (JET Contributors) 2021 Nucl. Fusion 61 106016) against the experimental results at JET, its reversed implementation is applied to derive the location of the modes. The experimental mode amplitude plays, this time, the role of the input data with the aim of conversely obtaining the perturbations location. The calculated location accuracy is conditioned by a good theoretical retrieval of the experimental mode amplitude and frequency. Based on the chosen initial conditions, the desired location is the one associated with the best possible mentioned retrieval. Our model reliability basically ensures the derivation of the suitable location. No safety factor and plasma rotational velocity data profiles are used. The method has been extensively tested and checked in order to become a valid alternative to the usual localization techniques.

Nuclear fusion could offer clean, abundant energy. However, managing the power exhausted from the core fusion plasma towards the reactor wall remains a major challenge. This is compounded in emerging compact reactor designs promising more cost-effective pathways towards commercial fusion energy. Alternative Divertor Configurations (ADCs) are a potential solution. In this work, we demonstrate exhaust control in ADCs, employing a novel method to diagnose the neutral gas buffer, which shields the target. Our work on the Mega Ampere Spherical Tokamak Upgrade shows that ADCs tackle key risks and uncertainties for fusion energy. Their highly reduced sensitivity to perturbations enables active exhaust control in otherwise unfeasible situations and facilitates an increased passive absorption of transients, which would otherwise damage the divertor. We observe a strong decoupling of each divertor from other reactor regions, enabling near-independent control of the divertors and core plasma. Our work showcases the real-world benefits of ADCs for effective heat load management in fusion power reactors.

The development of a high current baseline scenario ( I p = 3.5 M A , q 95 ≈ 3.0 , β N < 2 ) in deuterium (D), tritium (T) and deuterium-tritium (D-T) for high D-T fusion performance at JET with Be/W wall is described. We show that a suitable scenario capable of delivering up to 10 MW of fusion power, depending on the auxiliary heating power available, was successfully developed in D. However, when translated to T and D-T, the same scenario could not be sustained for the target duration of 5 s due to the impossibility to achieve the stationary compound edge localized modes regime necessary to flush the tungsten (W) from the plasma and control the density. Nevertheless, a peak fusion power in the order of 8 MW, with 60% of the power coming from thermal fusion reactions, was obtained in D-T at 3.5 MA, with ≈ 30 MW of neutral beam injection heating and 3-4 MW of ion cyclotron resonance heating, in line with the predictions obtained with the JINTRAC integrated scenario modelling suite of codes equipped with the QuaLiKiZ transport model and based on the extrapolation of the performance of similar D plasmas.

Exhausting power from the hot fusion core to the plasma-facing components is one fusion energy’s biggest challenges. The MAST Upgrade tokamak uniquely integrates strong containment of neutrals within the exhaust area (divertor) with extreme divertor shaping capability. By systematically altering the divertor shape, this study shows the strongest evidence to date to our knowledge that long-legged divertors with a high magnetic field gradient (total flux expansion) deliver key power exhaust benefits without adversely impacting the hot fusion core. These benefits are already achieved with relatively modest geometry adjustments that are more feasible to integrate in reactor designs. Benefits include reduced target heat loads and improved access to, and stability of, a neutral gas buffer that ‘shields’ the target and enhances power exhaust (detachment). Analysis and model comparisons shows these benefits are obtained by combining multiple shaping aspects: long-legged divertors have expanded plasma-neutral interaction volume that drive reductions in particle and power loads, while total flux expansion enhances detachment access and stability. Containing the neutrals in the exhaust area with physical structures further augments these shaping benefits. These results demonstrate strategic variation in the divertor geometry and magnetic topology is a potential solution to one of fusion’s power exhaust challenge. (Figure presented.)

Changing the isotope mass from deuterium to tritium in JET-ILW leads to an increase in the electron pedestal pressure (peped) due to an increase in the electron pedestal density (neped) [1]. The increase of the pressure with increasing effective mass (Aeff) has been explained via resistive MHD and by a change in the diamagnetic stabilization [1]. The results discussed in [1] were however obtained with a simple model for the diamagnetic stabilization. The present work aims to more properly model these effects using the code JOREK [2], which self-consistently treats the diamagnetic flows. The analysis has started with the JET deuterium shot 96208 at 2MA/2.3T. The pedestal height has then been increased, by adding heat and particle sources at the pedestal top, until an unstable mode starts growing. When the Teped is increased at fixed neped, no unstable modes are found. However, when neped is increased at fixed Teped a ballooning mode is driven unstable by just a small increase in the density. This can be explained by the fact that the diamagnetic frequency, which scales as ω* ∝ ∇p/n, also increases with the pressure gradient drive when Teped is increased but when neped is increased there is a negligible change in the diamagnetic frequency. This shows that the balance between the diamagnetic stabilization and the instability drive is the key to understanding the ELM-triggering mechanism in this plasma (and not just the instability drive itself). This has also been noted in previous studies of type-I ELMy pedestals [3]. To assess the effect of the isotope mass, Aeff has been changed from 2 to 3. The increase in Aeff led to a higher critical neped. This increase in neped is comparable to the increase in neped that is achieved by only rescaling the diamagnetic flow terms consistent with a change from Aeff = 2 to 3. We can therefore conclude that it is the change in the diamagnetic flows that increases neped when we go from deuterium to tritium. Quantitatively, the predicted increase in neped is smaller than the one seen in experiment, but qualitatively a good agreement is obtained.

References[1] L. Frassinetti et al 2023 Nucl. Fusion 63 112009[2] M. Hoelzl et al 2021 Nucl. Fusion 61 065001[3] A. Cathey et al 2020 Nucl. Fusion 60 124007

Turbulent transport is a decisive factor in determining the pedestal structure of H-modes. Here, we present the first comprehensive characterization of gyrokinetic turbulent transport in a JET hybrid H-mode pedestal. Local, linear simulations are performed to identify instabilities and global, nonlinear electromagnetic simulations reveal the turbulent heat and particle flux structure of the pedestal. Our analysis focuses on the Deuterium reference discharge #97781 performed in the scenario development for the Deuterium-Tritium campaign. We find the pedestal top transport to be dominated by ion temperature gradient (ITG) modes. In the pedestal center turbulent ion transport is suppressed and electron transport is driven by multi-faceted electron temperature gradient (ETG) modes, which extend down to ion-gyroradius scales. E × B shear is observed to strongly reduce the absolute turbulence level in global, nonlinear simulations. Furthermore, impurities are shown to reduce the main ion transport. Dedicated density and ion temperature profile variations test the sensitivity of the results and do not find strong differences in the turbulent transport in more reactor-like conditions.

This work examines the separatrix and Scrape-off-Layer (SOL) characteristics in three scenarios on JET: the Quasi-Continuous Exhaust (QCE) regime, the ITER Baseline scenario, and the X-point Radiator (XPR) regime. All three scenarios are potentially compatible with reactor operations, as they aim to provide power exhaust solutions through different approaches. The QCE regime is distinguished by its generally higher separatrix and SOL collisionality, associating with broader SOL width. These features, combined with the near-double-null (DNX) configuration, introduce several operational challenges on JET. The resulting broader SOL interacts with fast Beam neutrals, contributing to an unfavorable power load on local limiter. The heat load on the Upper Dump Plate Tiles in the QCE regime can be up to 5-6 times higher compared to the other scenarios. Additionally, the energy distribution shows a pronounced inner-outer asymmetry in QCE pulses, with the energy deposited on the outer limiter being up to four times higher than on the inner limiter. However, through careful operational planning and robust real-time protection system, the power loads were effectively managed within acceptable limits during QCE pulses, enabling successful scientific outcomes. As a result, the QCE regime serves as a case study to illustrate the critical need for integrating physics understanding, risk identification, operational strategies, and robust real-time protection to successfully implement new scenarios for fusion devices.

The effects of a poloidally localized bulge at outer midplane are investigated with regards to the stability of ideal MHD peeling-ballooning modes (PBMs) and n = ∞ ballooning modes. The effect of the bulge is that the n = ∞ ballooning mode stability at the bottom of the H-mode pedestal is significantly degraded, while the PBM stability is only slightly affected. The ratio of separatrix to pedestal density at which the n = ∞ ballooning mode limit is reached before the peeling-ballooning limit, decreases with the bulge size. The reduction of the required separatrix density would extend the access to quasi-continuous exhaust (QCE) operating mode without large edge localised modes to a wide range of scrape-off layer conditions. The free boundary equilibrium calculation shows that a bulge required for a significant stability change for a Spherical Tokamak for Energy Production fusion reactor can be achieved with 600 kA current in a midplane shaping coil.

Pedestals limited by peeling instabilities have been reached in JET-ILW by operating at high q95, up to q95=8.5. The increase in q95 via the increase of the toroidal field has stabilized the ballooning modes and has allowed to reach high pedestal temperature (up to 1.5 keV for the electrons and up to 2.2 keV for the ions) and low pedestal density ( approximate to 1.8x1019m-3), with electron-electron pedestal collisionality approximately 0.15 and normalized ion Larmor radius 0.002 approaching ITER normalized pedestal parameters. The most unstable pedestal instabilities are peeling with toroidal mode numbers in the range n=1-5. A density scan in peeling limited pedestals shows that the increase of the pedestal density leads to an increase in the pedestal pressure. The modeling shows that this effect is due to the stabilizing effect of the density on the peeling modes. On the contrary, the increase of the separatrix density does not seem to affect the pedestal pressure in peeling limited plasmas. These behaviors are opposite to those observed in ballooning limited pedestals. An isotope mass scan from pure deuterium to tritium-rich plasmas has been performed with peeling limited pedestals. The increase of the isotope mass leads to an increase of the density at the pedestal top, via the increase of the density gradient. This behavior is similar to that observed in ballooning limited pedestals. The increase of the isotope mass also leads to the increase of the pressure at the pedestal top, via the increase in the pressure gradient. The temperature is not affected significantly. The increase in the pressure is not ascribed to a direct effect of the isotope mass on the pedestal stability, but to an indirect effect due to the increase of the pedestal density which, as shown in the deuterium density scan, stabilizes the peeling modes. The experimental results are used to validate the pedestal predictions using the Europed code. In all the scans performed, a good qualitative agreement is observed between the predictions and the experimental results. Quantitative disagreements can be in part ascribed to the fact that a consistent modeling should integrate the effect of core and scrape-off layer.

Magnetohydrodynamic (MHD) stability simulations are central to predicting the performance of pedestals in high-confinement mode plasmas. A machine learning surrogate model, called KARHU, for the ideal MHD stability code MISHKA has been developed using a feed-forward convolutional neural network trained on a database of equilibrium simulations spanning a subset of the JET-ILW parameter space. A dataset of about 16 000 equilibria was created and MISHKA was used to assess the stability of these equilibria for eight toroidal mode numbers ranging between 3 and 50. KARHU was then trained to predict the maximum growth rate out of these toroidal mode numbers. The surrogate model was integrated into the Europed workflow. The Europed predictions using the surrogate model were compared to respective predictions using Europed with MISHKA, in order to demonstrate the improvement in simulation time and the accuracy of the predictions. A Europed run for an example scan was accelerated by 72%, where the MHD stability evaluation part of the model took less than 1% of the runtime. The accuracy was not compromised significantly. While the equilibria in this proof-of-principle work assume the standard Europed ballooning critical profile constraint to reduce the dimensionality of the dataset, the surrogate model was also tested on equilibria outside this constraint. Even for these equilibria that are strictly speaking outside the training domain, the model retains relatively good prediction performance within an average error of 22% for these pressure profiles.