The work describes the pedestal structure, transport and stability in an effective mass (A eff) scan from pure deuterium to pure tritium plasmas using a type I ELMy H-mode dataset in which key parameters that affect the pedestal behaviour (normalized pressure, ratio of the separatrix density to the pedestal density, pedestal ion Larmor radius, pedestal collisionality and rotation) are kept as constant as possible. Experimental results show a significant increase of the density at the pedestal top with increasing A eff, a modest reduction in the temperature and an increase in the pressure. The variations in the pedestal heights are mainly due to a change in the pedestal gradients while only small differences are observed in the pedestal width. A clear increase in the pedestal density and pressure gradients are observed from deuterium to tritium. The experimental results suggest a reduction of the pedestal inter-edge localized mode (inter-ELM) transport from deuterium to tritium. The reduction is likely in the pedestal inter-ELM particle transport, as suggested by the clear increase of the pedestal density gradients. The experimental results suggest also a possible reduction of the pedestal inter-ELM heat transport, however, the large experimental uncertainties do not allow conclusive claims on the heat diffusivity. The clear experimental reduction of eta e (the ratio between density and temperature gradient lengths) in the middle/top of the pedestal with increasing A eff suggests that there may be a link between increasing A eff and the reduction of electron scale turbulent transport. From the modelling point of view, an initial characterization of the behaviour of pedestal microinstabilities shows that the tritium plasma is characterized by growth rates lower than the deuterium plasmas. The pedestal stability of peeling-ballooning modes is assessed with both ideal and resistive magnetohydrodynamics (MHD). No significant effect of the isotope mass on the pedestal stability is observed using ideal MHD. Instead, resistive MHD shows a clear increase of the stability with increasing isotope mass. The resistive MHD results are in reasonable agreement with the experimental results of the normalized pedestal pressure gradient. The experimental and modelling results suggest that the main candidates to explain the change in the pedestal are a reduction in the inter-ELM transport and an improvement of the pedestal stability from deuterium to tritium.

Beam-target reactions are responsible for a substantial fraction of the fusion power generated in D-T plasmas in JET-ILW (Be/W-wall), with ion temperatures of 10-12keV and large neutral-beam injection (NBI) power. It is known that injecting D beam ions with energies of ∼100-150keV in T-rich plasmas has a larger potential for beam-target fusion than in 50:50 D:T plasmas, but such a scenario was never developed in the past D-T experiments performed in JET-C (Carbon-wall) and in TFTR in the 90's. On top of the intrinsic advantages of using D beams in T-rich plasmas for D-T neutron production, simulations have shown that fundamental ion-cyclotron resonance heating (ICRH) of the D ions can significantly boost the net fusion reactivity, since both the thermalized D ions and the fast D-NBI ions are accelerated to energy ranges that are optimal for the D-T reaction cross-section. The beneficial effect of fundamental D ICRH on thermal D minorities in tritium plasmas (without NBI) was identified in the JET-C D-T experiments, but was not tested in high performance H-mode discharges with D-NBI heating. In 2021, dedicated JET-ILW DTE2 [1] experiments confirmed - for the first time - the improved fusion performance of T-rich plasmas with high D-NBI power and highlighted the key impact of fundamental D ICRH on the fusion reactivity. This new scenario lead to the world-wide 5s averaged fusion power (and energy) record in D-T tokamak plasmas with dominant beam-target reactions. A brief experimental overview followed by detailed RF wave / Fokker-Planck simulations including NBI-ICRH synergy will be presented, to disentangle the different components contributing to the high neutron yield achieved in these experiments.

A reduction of the pedestal pressure with increasing separatrix density over pedestal density (n (e) (sep)/n (e) (ped)) has been observed in JET. The physics behind this correlation is investigated. The correlation is due to two distinct mechanisms. The increase of n (e) (sep)/n (e) (ped) till approximate to 0.4 shifts the pedestal pressure radially outwards, decreasing the peeling-balloning stability and reducing the pressure height. The effect of the position saturates above n (e) (sep)/n (e) (ped) approximate to 0.4. For higher values, the reduction of the pedestal pressure is ascribed to increased turbulent transport and, likely, to resistive MHD effects. The increase of n (e) (sep)/n (e) (ped) above approximate to 0.4 reduces backward difference n (e) /n (e), increasing eta (e) and the pedestal turbulent transport. This reduces the pressure gradient and the pedestal temperature, producing an increase in the pedestal resistivity. The work suggests that the increase in resistivity might destabilize resistive balloning modes, further reducing the pedestal stability.

A global heat flux model based on a fractional derivative of plasma pressure is proposed for the heat transport in fusion plasmas. The degree of the fractional derivative of the heat flux, α, is defined through the power balance analysis of the steady state. The model was used to obtain the experimental values of α for a large database of the Joint European Torus (JET) carbon-wall as well as ITER like-wall plasmas. The fractional degrees of the electron heat flux are found to be α<2, for all the selected pulses in the database, suggesting a deviation from the diffusive paradigm. Moreover, the results show that as the volume integrated input power is increased, the fractional degree of the electron heat flux converges to α∼0.8, indicating a global scaling between the net heating and the pressure profile in the high-power JET plasmas. The model is expected to provide insight into the proper kinetic description for the fusion plasmas and improve the accuracy of the heat transport predictions.

A power-balance model, with radiation losses from impurities and neutrals, gives a unified description of the density limit (DL) of the stellarator, the L-mode tokamak, and the reversed field pinch (RFP). The model predicts a Sudo-like scaling for the stellarator, a Greenwald- like scaling, alpha I-p(8/9), for the RFP and the ohmic tokamak, a mixed scaling, alpha (PIp4/9)-I-4/9, for the additionally heated L-mode tokamak. In a previous paper (Zanca et al 2017 Nucl. Fusion 57 056010) the model was compared with ohmic tokamak, RFP and stellarator experiments. Here, we address the issue of the DL dependence on heating power in the L-mode tokamak. Experimental data from high-density disrupted L-mode discharges performed at JET, as well as in other machines, arc taken as a term of comparison. The model fits the observed maximum densities better than the pure Greenwald limit.

Block-structured mesh generation techniques have been well addressed in the CFD community for automobile and aerospace studies, and their applicability to magnetic fusion is highly relevant, due to the complexity of the plasma-facing wall structures inside a tokamak device. Typically applied to non-linear simulations of MHD instabilities relevant to magnetically confined fusion, the JOREK code was originally developed with a 2D grid composed of isoparametric bi-cubic Bezier finite elements, that are aligned to the magnetic equilibrium of tokamak plasmas (the third dimension being represented by Fourier harmonics). To improve the applicability of these simulations, the grid-generator has been generalised to provide a robust extension method, using a block-structured mesh approach, which allows the simulations of arbitrary domains of tokamak vacuum vessels. Such boundary-aligned grids require the adaptation of boundary conditions along the edge of the new domain. Demonstrative non-linear simulations of plasma edge instabilities are presented to validate the robustness of the new grid, and future potential physics applications for tokamak plasmas are discussed. The methods presented here may be of interest to the wider community, beyond tokamak physics, wherever imposing arbitrary boundaries to quadrilateral finite elements is required.

The impurity concentration in the tokamak divertor plasma is a necessary input for predictive scaling of divertor detachment, however direct measurements from existing tokamaks in different divertor plasma conditions are limited. To address this, we have applied a recently developed spectroscopic N II line ratio technique for measuring the N concentration in the divertor to a range of H-mode and L-mode plasma from the ASDEX Upgrade and JET tokamaks, respectively. The results from both devices show that as the power crossing the separatrix, P-sep, is increased under otherwise similar core conditions (e.g. density), a higher N concentration is required to achieve the same detachment state. For example, the N concentrations at the start of detachment increase from approximate to 2% to approximate to 9% as P-sep, is increased from approximate to 2.5 MW to approximate to 7 MW. These results tentatively agree with scaling law predictions (e.g. Goldston et al.) motivating a further study examining the parameters which affect the N concentration required to reach detachment. Finally, the N concentrations from spectroscopy and the ratio of D and N gas valve fluxes agree within experimental uncertainty only when the vessel surfaces are fully-loaded with N.

Inconel-600 blocks and stainless steel covers for quartz microbalance crystals from remote corners in the JET-ILW divertor were studied with time-of-flight elastic recoil detection analysis and nuclear reaction analysis to obtain information about the areal densities and depth profiles of elements present in deposited material layers. Surface morphology and the composition of dust particles were examined with scanning electron microscopy and energy-dispersive X-ray spectroscopy. The analyzed components were present in JET during three ITER-like wall campaigns between 2010 and 2017. Deposited layers had a stratified structure, primarily made up of beryllium, carbon and oxygen with varying atomic fractions of deuterium, up to more than 20%. The range of carbon transport from the ribs of the divertor carrier was limited to a few centimeters, and carbon/deuterium co-deposition was indicated on the Inconel blocks. High atomic fractions of deuterium were also found in almost carbon-free layers on the quartz microbalance covers. Layer thicknesses up to more than 1 micrometer were indicated, but typical values were on the order of a few hundred nanometers. Chromium, iron and nickel fractions were less than or around 1% at layer surfaces while increasing close to the layer-substrate interface. The tungsten fraction depended on the proximity of the plasma strike point to the divertor corners. Particles of tungsten, molybdenum and copper with sizes less than or around 1 micrometer were found. Nitrogen, argon and neon were present after plasma edge cooling and disruption mitigation. Oxygen-18 was found on component surfaces after injection, indicating in-vessel oxidation. Compensation of elastic recoil detection data for detection efficiency and ion-induced release of deuterium during the measurement gave quantitative agreement with nuclear reaction analysis, which strengthens the validity of the results.

Hot spots on the divertor tiles at JET result in overestimation of the tile surface temperature which causes unnecessary termination of pulses. However, the appearance of hot spots can also indicate the condition of the divertor tile surfaces. To analyse the behaviour of the hot spots in the outer divertor tiles of JET, a simple image processing algorithm is developed. The algorithm isolates areas of bright pixels in the camera image and compares them to previously identified hot spots. The activity of the hot spots is then linked to values of other signals and parameters in the same time intervals. The operation of the detection algorithm was studied in a limited pulse range with high hot spot activity on the divertor tiles 5, 6 and 7. This allowed us to optimise the values of the controlling parameters. Then, the wider applicability of the method has been demonstrated by the analysis of the hot spot behaviour in a whole experimental campaign.

The analysis of the polarimetry measurements has the aim of validating models (De Marco and Segre 1972 Plasma Phys. 14 245), with a careful attention to the clarification of their limits of application. In this paper a new approximation method is introduced, the so-called special constant Omega direction (SCOD), which gives an analytical solution to the polarimetry exact Stokes model equations. The available approximate solutions (including SCOD) of the polarimetry propagation equations are presented, compared and their application limits determined, using a reference tokamak configuration, which is a simplified equilibrium for a circular tokamak. The SCOD approximation is compared successfully to the Stokes model in the context also of equilibria evaluated for two JET discharges. The approximation methods are analytical or very simple mathematical expressions which can also be used in equilibrium codes for their optimization.