Pedro Carvalho

Screening of high-Z (W) impurities from the confined plasma by the temperature gradient at the plasma periphery of fusion-grade H-mode plasmas was demonstrated for the first time in JET with the Be/W wall [Field A R et al , Nucl. Fusion 63(2023) 016028]. Additional experiments have been performed in JET during 2023, including in DT during the DTE3 campaign, to further optimise the impurity screening in such plasmas, as well as our bolometric measurements of the W impurity fluxes between and during edge-localised modes (ELMs). A decrease in plasma current from 2.3MA to 2.1 MA, reduced the electron density and thereby increased the ion temperature at the H-mode pedestal top, resulting in stronger impurity screening behaviour. The scenario was then successfully transferred to operation in DT by increasing the toroidal field, in order to compensate the lower L/H-threshold power in DT compared to D plasmas. Here, results of detailed analysis and modelling of the neo-classical (NC) W transport infour pulses from these experiments are presented, two in D at 2.3MA and 2.1MA plasma current and a matched pulse pair at 2.1MA in D and DT. Using the FACIT code [D Fajardo et al 2023 Plasma Phys. Control. Fusion 65 035021] to model the NC W transport for these more recent pulses, the outward convection just inside the pedestal top found in our earlier study could not be reproduced. Possible reasons for this discrepancy between experimental observations and our modeling results are discussed, including potential deficiencies in our measurement technique and/or incompleteness of the NC transport modelling.

The empirical scaling for the H-mode power threshold in tokamaks has been revisited using a database with threshold data from machines with a metallic first wall as part of International Tokamak Physics Activity (ITPA) task TC-26. The database contains discharges from ASDEX Upgrade (W), JET (Be/W) and Alcator C-Mod (Mo). This was motivated by reports that in like-for-like discharges the power threshold was reduced by approximately 30% after the change from carbon based to metallic first wall materials in AUG [F. Ryter et al., Nuclear Fusion, 53(11):113003 (2013)] and JET [C.F. Maggi et al., Nuclear Fusion, 54(2):023007 (2014)]. The database contains L-H transition data for all hydrogen isotopes and mixtures, including T and DT from the recent JET campaigns. compared to the ITPA 2008 scaling [Y R Martin et al., Journal of Physics: Conference Series, 123(1):012033 (2008)], the metal wall scaling has a smaller magnetic field exponent but a larger density exponent. We present an additional parameter to capture the strong dependence of the L-H power threshold (approx. factor 2) on the magnetic configuration in the divertor on JET. The scaling recovers the approximate inverse isotope mass scaling of the threshold power. Alternative scalings involving the plasma current and poloidal magnetic field are explored. Despite the reduction in threshold observed earlier, the scalings based on the metal wall database do not necessarily extrapolate to a lower threshold for ITER compared to the ITPA 2008 scaling, especially at high density. The divertor configuration effect induces the largest uncertainty in the extrapolation.

Nonlinear 3D magnetohydrodynamic (MHD) modeling of shattered pellet injection (SPI) in JET tokamak plasmas is performed with the JOREK code. The study focuses on the validation of simulation result with respect to experimental observations, addressing in particular figures of merit for the efficiency of the SPI, as a technique to mitigate thermal loads by injecting radiative impurities like neon during the thermal quench (TQ) phase of disruptions. A set of JET pulses with neon/deuterium atomic mixture ratio in the shattered pellet varying from 10% to 100% provides the experimental data to be compared with. Simulations using different models for the ablation of solid fragments and the radiation of neon impurities are considered. Synthetic diagnostics are employed for a direct quantitative comparison with key JET diagnostic systems providing in particular radiation, electron temperature and density, and magnetic measurements. 3D radiation structures from MHD simulations are analyzed to estimate the toroidal asymmetry of radiation, which is difficult to measure in JET due to the presence of only two toroidally displaced multi-channel bolometry systems. JOREK simulations show that, before the TQ, the bulk of the radiation is concentrated in a toroidal region enclosing the radiating fragments that is missed by both multi-channel bolometers, suggesting a possible underestimation of the total radiation by such diagnostic systems in the pre-TQ phase. On the other hand, JOREK predicts that during the TQ the radiation becomes more toroidally symmetric, with positive implications for the heat load to plasma facing components.

JET returned to deuterium-tritium operations in 2023 (DTE3 campaign), approximately two years after DTE2. DTE3 was designed as an extension of JET’s 2022-2023 deuterium campaigns, which focused on developing scenarios for ITER and DEMO, integrating in-depth physics understanding and control schemes. These scenarios were evaluated with mixed D-T fuel, using the only remaining tritium-capable tokamak until its closure in 2023. A core-edge-SOL integrated H-mode scenario was developed and tested in D-T, showing good confinement and partial divertor detachment with Ne-seeding. Stationary pulses with good performance, no tungsten accumulation, and even without ELMs were achieved in D-T. Plasmas with pedestals limited by peeling modes were studied with D, T-rich, and D-T fuel, revealing a positive correlation between pedestal electron pressure and pedestal electron density. The Quasi-Continuous Exhaust regime was successfully achieved with D-T fuel, with access criteria similar to those in D plasmas. A scenario with full detachment, the X-point radiator regime, was established in D-T, aided by the real-time control of the radiator’s position. The crucial characterisation of tritium retention continued in DTE3, using gas balance measurements and the new LID-QMS diagnostic. Nuclear technology studies were advanced during the DTE3 campaign, addressing issues such as the activation of water in cooling loops and single event effects on electronics. Building on the previous D, T and DTE2 campaigns and the lessons learned from them, DTE3 extended our understanding of D-T plasmas, particularly in scenarios relevant to next-generation devices such as ITER and DEMO.

Previous investigations on JET suggest half or less of plasma stored thermal energy Wth is radiated (frad,th≲0.5) using either massive gas injection (MGI) or shattered pellet injection (SPI) disruption mitigation. We investigate whether the apparent incomplete frad,th is explained by radiation peaking near the injection plume. High toroidal peaking throughout the pre-thermal quench is found in argon–deuterium MGI on JET, with typically >3× higher radiation near the injector than toroidally distant. Previously unexplained toroidal bolometry measurements in neon–deuterium SPI are reproduced with similar peaking using the Emis3D radiation analysis code. These observations align with results from Alcator C-Mod and KSTAR. This peaking is not captured by previous JET studies that found poor thermal mitigation. Two sets of neon–deuterium SPI and two sets of argon–deuterium MGI are analyzed using Emis3D. In SPI, frad,th rises from no-plume estimates of 0.31 and 0.66 to lower bounds of 0.84 and 0.92, respectively, and frad,th∼1 is possible. In MGI, the toroidal spread of the peaking feature is poorly constrained. frad,th up to 0.85 and 0.65 are possible using the largest possible spread, increasing from 0.42 and 0.28, although frad,th∼1 does not appear to be reached. Revised mitigation estimates on JET suggest a lower melt risk to the divertor in mitigated disruptions on ITER and SPARC than previously thought. However, peaking near injectors could increase flash melting risk on nearby plasma facing components.

Components on the inner wall of fusion reactors are exposed to extremely high heat loads, and significant care must be taken to minimize the damage to these components during operation. UFOs, also known as transient impurity events, are small particles of dust within the tokamak vessel that can lead to plasma disruptions. These can cause serious damage to the device and remain a significant challenge for safe operations. This study presents a novel approach to track these UFO events by utilizing visual camera data from the Joint European Torus and a Convolutional Neural Network to identify UFOs within the tokamak. The model attained an accuracy of 95.1%, a precision of 95.4% for UFO detection, a recall of 95.1%, and a receiver operating characteristic curve area of 0.99. These metrics underscore the model's competence in reliably identifying UFOs with minimal false identifications. Consequently, this suggests the model's suitability for integration into UFO detection systems in tokamak environments, improving monitoring and efficiency, and potentially providing an avenue for real-time UFO detection in future research.

Massive material injections in the JET tokamak have been observed to substantially affect resistive bolometer measurements, resulting in a spurious radiated power signal proportional to the quantity injected and reaching up to 8 MW. These bolometers are calibrated and designed to operate in near vacuum but certain scenarios requiring large gas injections can push the neutral pressure past nominal values. This study demonstrates that the bolometry measurement can be affected at neutral pressures above 0.1 Pa following injections with standard gas valves, shattered pellet injections, and particularly massive gas injections. The power measurement of resistive bolometers is based on the temperature difference between a measurement sensor exposed to radiation and a shielded reference sensor. We employ a thermal conductivity model to demonstrate that the conduction through the gas and the distinct geometries between the sensors can affect their cooling efficiency. This additional cooling pathway, coupled with the Joule heating from the applied voltage causes the equilibrium temperatures of the sensors to diverge. Being the very basis of the measure, this temperature difference induces a signal that is erroneously interpreted as radiated power. Experiments show large discrepancies in the response to neutral pressure among bolometer channels, attributed to variations in channel physical parameters. Nonetheless, the modeled total radiated power reproduces the experimental measurements within an order of magnitude, affirming the sensitivity of resistive bolometers to neutral pressure and gas species.

Experimental campaigns at the Joint European Torus with an ITER-like Be/W wall with pure deuterium (D), tritium (T), and deuterium–tritium (DT) were a unique opportunity to explore various aspects related to the ITER operation. One of the most important challenges in recent years was the development of the hybrid scenario for D-T, based on reference deuterium and tritium plasmas. This kind of scenario, one of the foreseen for ITER, is characterized by a low current plasma and a high normalized beta βN factor compared to the parallel optimized baseline scenario [Hobirk et al., Plasma Phys. Controlled Fusion 54, 095001 (2012)]. As the experiments have shown, controlling the plasma edge in the different phases of the hybrid scenario becomes more difficult with higher isotope mass, and therefore, are also in risk of impurity accumulation [Hobirk et al., Nucl. Fusion 63, 112001 (2023)]. For this reason, investigation of the impurity behavior, as well as their control, constituted the crucial issue. The present contribution aims to compare mid-Z and high-Z impurities behavior within H-mode hybrid discharges in D and T plasmas, as well as D and DT plasmas. Detailed analysis shows that in the H-mode regime in the hybrid scenario, higher impurity radiation is observed for DT in comparison to D plasmas, as well as for T compared to D plasmas. Additionally, it was noticed that the most significant contribution to the plasma radiated power comes from W and to a lesser extent from Ni (∼10%). Moreover, it was found that an earlier transition from small edge localized modes (ELMs) to ELM-free phase can result in the earlier increase in impurities.