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.

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

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.

Fusion power is a promising candidate for providing large amounts of sustainable, planable power to complement other sustainable energy solutions in the future. The tokamak, which is the fusion device furthest along to achieving this goal, confines a hot plasma with the help of magnetic fields. The performance of the tokamak is highly dependent on the performance of a thin region near the plasma edge, called the pedestal. Accurate models for predicting the pedestal behavior is therefore paramount for the optimization of future fusion reactors. The pedestal height is typically limited by the onset of ideal magnetohydrodynamic (MHD) instabilities called edge localized modes (ELMs). In the JET tokamak, it has however been observed that sometimes the pre-ELM pedestal can sometimes be stable to ideal MHD modes.

This thesis investigates the physics which are required to reconcile modeling and experimental results in pedestals which are not marginally unstable to ideal MHD modes when the ELM is triggered. It is shown that a key component that seems to be missing is the lack of resistivity in the MHD modeling. To investigate the impact of resistivity on the MHD modeling, a resistive MHD code has been implemented into the MHD stability frameworks used at JET. Including the resistivity improves the agreement between model and experiment compared to ideal MHD. In particular, the impact of changing the main fuel isotope mass and the impurity content on the pedestal performance is captured when resistive MHD is used.

Fusion är en lovande källa av stora mängder hållbar och planerbar energi som kan kompletera andra hållbara energilösningar i framtiden. Tokamaken, som är det fusionskoncept som har kommit längst på vägen för att uppnå detta mål, innesluter ett varmt plasma med hjälp av magnetiska fält. Tokamakens prestanda är hårt bunden till prestandan hos en liten region nära kanten på plasmat som kallas pedestalen. För att kunna optimera framtida fusionsreaktorer så krävs pålitliga modeller för att förutspå pedestalens beteende. Höjden på pedestalen är oftast begränsad av ideala magnetohydrodynamiska (MHD) instabiliteter som kallas edge localized modes (ELMs). I tokamaken JET, så har det dock obeserverats att pedestalen ibland är stabil mot ideala MHD instabiliteter precis innan en ELM.

Den här avhandlingen undersöker vilken ytterligare fysik som krävs för att försona de teoretiska modellerna med de experimentella resultaten, när pedestalen inte är marginellt instabil mot ideala MHD moder när en ELM utlöses. Det visas att en nyckelkomponent är bristen på resistivitet i MHD modelleringen. För att undersöka effekten av att inkludera resistivitet på MHD modeleringen, så har en resistiv MHD kod blivit implementerad i MHD stabilitetsramverken som används på JET. Inkluderandet av resistivitet leder till att resultaten från simuleringarna stämmer bättre överens med de experimentella resultaten. Inkluderandet av resistivitet i modeleringen låter en även fånga effekten på pedestalen från en ändring av bränsleisotop och föroreningar i plasmat, vilket inte fångas av ideal MHD. 

The objective of thermonuclear fusion consists of producing electricity from the coalescence of light nuclei in high temperature plasmas. The most promising route to fusion envisages the confinement of such plasmas with magnetic fields, whose most studied configuration is the tokamak. Disruptions are catastrophic collapses affecting all tokamak devices and one of the main potential showstoppers on the route to a commercial reactor. In this work we report how, deploying innovative analysis methods on thousands of JET experiments covering the isotopic compositions from hydrogen to full tritium and including the major D-T campaign, the nature of the various forms of collapse is investigated in all phases of the discharges. An original approach to proximity detection has been developed, which allows determining both the probability of and the time interval remaining before an incoming disruption, with adaptive, from scratch, real time compatible techniques. The results indicate that physics based prediction and control tools can be developed, to deploy realistic strategies of disruption avoidance and prevention, meeting the requirements of the next generation of devices.

Tokamak à configuration variable (TCV), recently celebrating 30 years of near-continual operation, continues in its missions to advance outstanding key physics and operational scenario issues for ITER and the design of future power plants such as DEMO. The main machine heating systems and operational changes are first described. Then follow five sections: plasma scenarios. ITER Base-Line (IBL) discharges, triangularity studies together with X3 heating and N2 seeding. Edge localised mode suppression, with a high radiation region near the X-point is reported with N2 injection with and without divertor baffles in a snowflake configuration. Negative triangularity (NT) discharges attained record, albeit transient, βN ∼ 3 with lower turbulence, higher low-Z impurity transport, vertical stability and density limits and core transport better than the IBL. Positive triangularity L-Mode linear and saturated ohmic confinement confinement saturation, often-correlated with intrinsic toroidal rotation reversals, was probed for D, H and He working gases. H-mode confinement and pedestal studies were extended to low collisionality with electron cyclotron heating obtaining steady state electron iternal transport barrier with neutral beam heating (NBH), and NBH driven H-mode configurations with off-axis co-electron cyclotron current drive. Fast particle physics. The physics of disruptions, runaway electrons and fast ions (FIs) was developed using near-full current conversion at disruption with recombination thresholds characterised for impurity species (Ne, Ar, Kr). Different flushing gases (D2, H2) and pathways to trigger a benign disruption were explored. The 55 kV NBH II generated a rich Alfvénic spectrum modulating the FI fas ion loss detector signal. NT configurations showed less toroidal Alfvén excitation activity preferentially affecting higher FI pitch angles. Scrape-off layer and edge physics. gas puff imaging systems characterised turbulent plasma ejection for several advanced divertor configurations, including NT. Combined diagnostic array divertor state analysis in detachment conditions was compared to modelling revealing an importance for molecular processes. Divertor physics. Internal gas baffles diversified to include shorter/longer structures on the high and/or low field side to probe compressive efficiency. Divertor studies concentrated upon mitigating target power, facilitating detachment and increasing the radiated power fraction employing alternative divertor geometries, optimised X-point radiator regimes and long-legged configurations. Smaller-than-expected improvements with total flux expansion were better modelled when including parallel flows. Peak outer target heat flux reduction was achieved (>50%) for high flux-expansion geometries, maintaining core performance (H98 > 1). A reduction in target heat loads and facilitated detachment access at lower core densities is reported. Real-time control. TCV’s real-time control upgrades employed MIMO gas injector control of stable, robust, partial detachment and plasma β feedback control avoiding neoclassical tearing modes with plasma confinement changes. Machine-learning enhancements include trajectory tracking disruption proximity and avoidance as well as a first-of-its-kind reinforcement learning-based controller for the plasma equilibrium trained entirely on a free-boundary simulator. Finally, a short description of TCV’s immediate future plans will be given.

In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements-new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors-as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D-T operations is presented in (King et al Nucl. Fusion submitted).

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.