Designing a new tokamak requires concerted efforts of engineers and physicists. In order to reduce costs and minimise risks, a first-principle based integrated modelling as comprehensive as possible of plasma discharges in different operational scenarios is an essential tool. Therefore, main baseline scenarios of the future Divertor Tokamak Test facility (DTT) [1] (R0 = 2:19m, a = 0:70m,Wfirst wall and divertor, pulse length 100s, plasma current Ipl 5:5MA, vacuum toroidal field Btor 5:85T, total power by auxiliary heating systems Ptot 45MW) have been simulated extensively. This modelling work led to the optimisation of the device size and of the reference heating mix, as widely described in [2], and provided reference profiles for diagnostic system design, estimates of neutron yields, calculations of fast particle losses, gas puffing and/or pellet feature requirements for fuelling, MHD evaluations, and other tasks. The latest simulation results of the DTT scenarios with the Single Null magnetic configuration are presented here. These runs, carried out with the JINTRAC [3] suite or the ASTRA [4] transport solver, make use of theory based quasi-linear transport models (QLK [5] and TGLF SAT2 [6]), ensuring the highest fidelity presently achievable in integrated modelling. A specific attention to the consistency between the control coil system capabilities and plasma profiles has been paid and the edge requirements to have plasma scenarios compatible with divertor and first wall power handling capability and tungsten influx have been taken into account.

The scenario integrated modelling is a top priority work during the design of a new tokamak, as the Divertor Tokamak Test facility (DTT) under construction at the ENEA Research Center in Frascati. The first simulations of the main baseline scenarios contributed to the optimization of the DTT project, particularly with regard to the machine size and heating systems, besides serving as reference for diagnostics design. In this paper we report the first simulations of the full power baseline scenario in the final configuration of the machine and heating mix.

Previous studies with first-principle-based integrated modelling suggested that electron temperature gradient (ETG) turbulence may lead to an anti-gyroBohm isotope scaling in JET high-performance hybrid H-mode scenarios. A dedicated comparison study against higher-fidelity turbulence modelling invalidates this claim. Ion-scale turbulence with magnetic field perturbations included, can match the power balance fluxes within temperature gradient error margins. Multiscale gyrokinetic simulations from two distinct codes produce no significant ETG heat flux, demonstrating that simple rules-of-thumb are insufficient criteria for its onset.

The JET 2019-2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) installed in 2010, improved diagnostic capabilities now fully available, a major neutral beam injection upgrade providing record power in 2019-2020, and tested the technical and procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle (?) physics in the coming D-T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed shattered pellet injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILWplasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design and operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D-T benefited from the highest D-D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.

Designing a new tokamak requires concerted efforts of engineers and physicists. In order to reduce costs and minimise risks, a first-principle based integrated modelling as comprehensive as possible of plasma discharges in different operational scenarios is an essential tool. Therefore, main baseline scenarios of the future Divertor Tokamak Test facility (DTT) [1] (R0 = 2:19m, a = 0:70m,Wfirst wall and divertor, pulse length 100s, plasma current Ipl 5:5MA, vacuum toroidal field Btor 5:85T, total power by auxiliary heating systems Ptot 45MW) have been simulated extensively. This modelling work led to the optimisation of the device size and of the reference heating mix, as widely described in [2], and provided reference profiles for diagnostic system design, estimates of neutron yields, calculations of fast particle losses, gas puffing and/or pellet feature requirements for fuelling, MHD evaluations, and other tasks. The latest simulation results of the DTT scenarios with the Single Null magnetic configuration are presented here. These runs, carried out with the JINTRAC [3] suite or the ASTRA [4] transport solver, make use of theory based quasi-linear transport models (QLK [5] and TGLF SAT2 [6]), ensuring the highest fidelity presently achievable in integrated modelling. A specific attention to the consistency between the control coil system capabilities and plasma profiles has been paid and the edge requirements to have plasma scenarios compatible with divertor and first wall power handling capability and tungsten influx have been taken into account.

This is an overview of the theoretical understanding of the so-called isotope effect in JET hydrogen versus deuterium plasmas. Experimentally, weak to moderate deviations from naive GyroBohm scaling expectations are found for the core heat transport in L and H-modes. The physical mechanisms behind such deviations are analysed in the framework of the gyrokinetic theory. In the case of particle transport, isotope effects are mostly found in the plasma edge where the density is higher in deuterium than in hydrogen plasmas. In general, both the thermal energy and particle confinement increase with increasing main ion mass. A comparison of such results to expectations for deuterium-tritium plasmas in ITER is discussed.

New H-mode regimes with high confinement, low core impurity accumulation, and small edge-localized mode perturbations have been obtained in magnetically confined plasmas at the Joint European Torus tokamak. Such regimes are achieved by means of optimized particle fueling conditions at high input power, current, and magnetic field, which lead to a self-organized state with a strong increase in rotation and ion temperature and a decrease in the edge density. An interplay between core and edge plasma regions leads to reduced turbulence levels and outward impurity convection. These results pave the way to an attractive alternative to the standard plasmas considered for fusion energy generation in a tokamak with a metallic wall environment such as the ones expected in ITER.

An intensive integrated modelling work of the main scenarios of the new Divertor Tokamak Test (DTT) facility with a single null divertor configuration has been performed using first principle quasi-linear transport models, in support of the design of the device and of the definition of its scientific work programme. First results of this integrated modelling work on DTT (R0 = 2.14 m, a = 0.65 m) are presented here along with outcome of the gyrokinetic simulations used to validate the reduced models in the DTT range of parameters. As a result of this work, the heating mix has been defined, the size of device has been increased to R0 = 2.19 m and a = 0.70 m, the use of pellets for fuelling has been recommended and reference profiles for diagnostic design, estimates of neutron yields and fast particle losses have been made available.

Tokamaks dominated by electron heating like ITER could possibly suffer from the consequences of electron temperature gradient (ETG) mode destabilisation, which could develop a turbulent electron heat flux capable of setting an upper limit to the achievable electron temperature peaking, resulting in a degradation of the fusion performance. An effort is carried out in this paper to collect and compare the results of dedicated plasma discharges performed during the last few years at three of the major European tokamaks, TCV, AUG and JET, by analysing the electron heat transport for cases presumably compatible with ETGs relevance, given the actual theoretical understanding of these instabilities. The response of the electron temperature profiles to electron heat flux changes is experimentally investigated by performing both steady state heat flux scans and perturbative analysis by radio frequency heating modulation. The experimental results are confronted with numerical simulations, ranging from simple linear gyrokinetic (GK) or quasi-linear runs, to very computationally expensive nonlinear multi-scale GK simulations, resolving ion and electron scales at the same time. The results collected so far tend to confirm the previously emerging picture, indicating that cases with a proper balance of electron and ion heating, with similar electron and ion temperatures and sufficiently large ETG, could be compatible with a non negligible impact of ETGs on the electron heat transport. The ion heating destabilises ETGs not only by increasing the ion temperature but also thanks to the stabilisation of ion-scale turbulence by a synergy of fast ions and E × B shearing, which are in some cases associated with it. The stabilising effect of plasma impurities on ETGs is still under investigation by means of multi-scale GK simulations, and also direct experimental measurements of density and temperature fluctuations at electron scales would be needed to ultimately assess the impact of ETGs.

Dedicated electron heat transport experiments have been carried out in L- and H-mode Deuterium plasmas of the JET-ILWtokamak to identify the amount of electron heat carried by electron-scale electron temperature gradient (ETG) modes. Ion cyclotron resonance heating at different positions has been used to probe the response of the electron temperature inverse gradient length R/LTe to changes in electron heat flux qe, while different amounts of neutral beam heating allowed to scan the ratio of ion to electron temperature Te/Ti, which is a key parameter for the onset of ETGs. Results indicate a steepening of the normalized qe vs R/LTe curve above R/LTe ~ 8 for Te/Ti _ 1, suggestive of the ETG onset. Ion-scale gyro-kinetic (GK) simulations match the ion heat flux and the low-R/LTe part of the qe curve, but do not reproduce such steepening at high R/LTe. Multi-scale GK simulations covering both ion and electron scales and including one impurity bundling light and heavy species indicate an ETG contribution only for R/LTe values larger than the experimental ones. Sensitivity studies of such result are difficult to achieve due to limitation in numerical resources. The quasi-linear TGLF model has been used for sensitivity studies. With the same bundled impurity as the GK multi-scale, TGLF shows the qe steepening at much larger R/LTe values than in experiment, but when using the real mix of light impurities neglecting the heavy impurities, TGLF gets closer to the experimental results. Profile simulations with TGLF including both light and heavy impurities show over-prediction of Te profiles and in some cases also of density, but good Ti predictions, confirming issues with the model electron stiffness for these plasmas EURATOM 2021.

In the European Roadmap towards thermonuclear fusion power production, studying the controlled exhaust of energy and particles from a fusion reactor is a top priority research item. This is the main goal of the Divertor Tokamak Test (DTT) facility, a D-shaped superconducting tokamak (R = 2.19 m, a = 0.70 m, BT <= 6 T, Ip <= 5.5 MA, pulse length <= 100 s, auxiliary heating <= 45 MW, W first wall and divertor), whose construction is starting in Frascati. In order to support the device design and to help the elaboration of a DTT scientific work-programme, it is a key priority to achieve multi-channel integrated modelling of DTT scenarios based on state-of-art first-principle quasi-linear transport models. First modelling results of the main DTT scenarios are presented here. Steady-state profiles of ion and electron temperatures, densities, rotation, and current density were predicted with a calculated self-consistent equilibrium, with turbulent heat and particle transport calculated by the TGLF or QLK transport models, and with heating modelled self-consistently. As a result of this work, the heating mix was defined and reference profiles have been become available.

Up to 15s H-mode plasmas with 8MW of ICRH power were executed on JET, with the JET record high injected ICRH energy of 108MJ. The ICRH discharges, using the H minority heating scheme at 3% minority concentration, were stationary without any major MHD activities or impurity accumulation. NBI heated discharges were executed consecutively to match the dimensionless plasma profiles of q, ?*, ?*, ?n and Ti/Te using the same shape. Both pulses had gas puff modulation at 3Hz for the whole duration of the discharges to extract perturbative particle transport coefficients. The ELM frequencies are 75Hz and 45Hz for the ICRH and NBI pulses, respectively, with mixed ELMs and affected by gas puff modulation.

We propose a novel interpretation of the Edge Harmonic Oscillation (EHO), which is seen as part of quiescent high-confinement (QH) mode, and compare its predictions to JET data. This new model for the EHOs will improve our ability to predict the QH state. Unlike the standard high confinement (Hmode) regime, QH plasmas avoid violent and periodic plasma eruptions known as Edge Localised Modes (ELMs) [1]. ELMs deposit unacceptable peak heat loads causing a severe deterioration of the plasma facing components. This poses a serious operational threat to reactor relevant plasma scenarios. Thus, the development of naturally ELM-free regimes has become of crucial importance [2]. One such regime is the so called QH mode. In QH plasmas, ELMs are replaced by dominantly low- steady mild MHD EHOs [3]. EHOs have been observed in DIII-D [3], ASDEX-U [4], JET [5], JT60 [6] and NSTX [7] at low edge collisionality over a fairly broad range in [3]. EHOs, which are typically characterised by multiple rotating toroidal harmonics localised in the pedestal region [8], enhance particle transport allowing density control and potentially ash removal without the impulsive heat load caused by ELMs [9]. A successful access to and control of this favourable regime, still requires a deeper understanding of its mechanisms in terms of basic stability concept. We thus propose, within the ideal MHD linear stability framework, a novel interpretation for the EHO onset mechanism. Provided by the possibility of edge infernal-type instabilities in QH plasmas [10], we show that the interplay of poloidal flows (MHD and diamagnetic) allows the suppression of short wavelength modes so that low- oscillations, namely EHOs, can emerge [11]. Our model retrieves several features measured experimentally such as mode rotation frequencies, radial struture [8], and amplitude of the critical shearing rate [12]. The theoretical understanding of the EHO mode onset is then applied to the interpretation of JET-C discharges exhibiting Outer Mode activity [5] which are subsequently compared with the JET-ILW database. [1] A. W. Leonard, Phys. Plasmas 21, 090501 (2014); [2] E. Viezzer et al., Nucl. Fusion 58, 115002 (2018); [3] K. H. Burrell et al., Phys. Plasmas 12, 056121 (2005); [4] W. Suttrop et al., Plasma Phys. Control. Fusion 45, 1399 (2003); [5] E. R. Solano et al., Phys. Rev. Lett. 104, 185003 (2010); [6] N. Oyama et al., Nucl. Fusion 45, 871 (2005); [7] K. F. Gan et al., Nucl. Fusion 57, 126053 (2017); [8] X. Chen et al., Nucl. Fusion 56, 076011 (2016); [9] K. H. Burrell et al., Phys. Rev. Lett. 102, 155003 (2009); [10] D. Brunetti et al., Nucl. Fusion 58, 014002 (2018); [11] D. Brunetti et al., Phys. Rev. Lett. 122, 155003 (2019); [12] T. M. Wilks et al., Nucl. Fusion 58, 112002 (2018)

ccurate predictive modelling of tokamak core turbulent transport is a vital component of in- tegrated tokamak simulation. The contribution of Electron Temperature Gradient (ETG) driven turbulence to electron heat transport in various operational regimes is an open question with extensive recent investigations [1], [2]. This paper focuses on validation of striking recent pre- dictions [3] of anti-GyroBohm isotope scaling of core transport, mediated by ETG turbulence.

Tokamaks dominated by electron heating like ITER could possibly suffer from the consequences of an elec-tron temperature gradient (ETG) mode destabilisation, which could develop a turbulent electron heat flux ca-pable of setting an upper limit to the achievable electron temperature peaking, resulting in a degradation of thefusion performances. An effort is carried out in the paper to collect and compare the results of dedicated plasmadischarges performed during the last years at three of the major European tokamaks, TCV, AUG and JET, byanalysing the electron heat transport for cases presumably compatible with ETGs relevance given the actual the-oretical understanding of these instabilities. The response of the electron temperature profiles to electron heatflux changes is experimentally investigated by performing both steady state heat flux scans and perturbativeanalysis by radio frequency modulation. The experimental results are confronted with numerical simulations,ranging from simpler linear gyrokinetic or quasi-linear runs, to very computationally expensive nonlinear multi-scale gyrokinetic simulations, resolving ion and electron scales at the same time. The results collected so fartend to confirm the previously emerging picture indicating that only cases with a proper balance of electron andion heating, with similar electron and ion temperatures and sufficiently large electron temperature gradient, arecompatible with a non negligible impact of ETGs on the electron heat transport. The ion heating destabilisesETGs not only by increasing the ion temperature but also thanks to the stabilisation of ion-scale turbulence bya synergy of fast ions andE×Bshearing which are in some cases associated to it. The stabilising effect ofplasma impurities on ETGs is still under investigation by means of multi-scale gyrokinetic simulations, and alsodirect experimental measurements of density and temperature fluctuations at electron scales would be needed toultimately assess the impact of ETGs.

A benchmark of the reduced quasi-linear models QuaLiKiz and TGLF with GENE gyrokinetic simulations has been performed for parameters corresponding to a JET high performance hybrid pulse in deuterium. Given the importance of the study of such advanced scenarios in view of ITER and DEMO operations, the dependence of the transport on the ion isotope mass has also been assessed, by repeating the benchmark changing the ion isotope to tritium. TGLF agrees better with GENE on the linear spectra and the flux levels. However, concerning the isotope dependence, only QuaLiKiz reproduces the GENE radial trend of a basically gyro-Bohm (gB) scaling at inner radii and instead anti-gB at outer radii. The physics effects which are responsible of the antigB effect in GENE simulations have been singled out.

The Divertor Tokamak Test facility (DTT) [1-3] is a D-shaped superconducting tokamak (R=2.14 m, a=0.65 m, BT<=6 T, Ip<= 5.5 MA, pulse length <= 100 s, auxiliary heating <= 45 MW, W first wall and divertor), whose construction is starting in Frascati, Italy. Its main mission is to study the controlled exhaust of energy and particle from a fusion reactor, which is a top priority research item in the European Roadmap [4] towards thermonuclear fusion power production. This will be possible in DTT by achieving large PSEP/R values (where PSEP is the power flowing through the last closed magnetic surface) using 45 MW of auxiliary heating in a high performance machine characterised by high flexibility in the choice of the divertor and of the magnetic configurations. The characteristics of the machine will allow to address many ITER and DEMO relevant physics issues besides plasma wall interaction in a fusion relevant range of plasma parameters. The heating mix foresees the use of 170 GHz ECRH, 60-90 MHz ICRH and 400 keV negative ion beam injectors, with ECRH being the main system, although the precise sharing between the three systems has still to be optimised. In order to help with the heating system definition, and to provide scenarios for the design of diagnostics and pellet injector, or for the evaluation of issues such as ripple losses or neutron shields, it is a key priority to achieve multi-channel integrated modelling of DTT scenarios based on state-of-art first principle quasi-linear transport models, whose reliability stems from an extensive validation work against experiments and high fidelity gyrokinetic simulations carried out within the EUROfusion and ITPA frameworks (see e.g. the recent overview [5] and references therein). It is also important that the integrated modelling results for some cases are validated against gyrokinetic simulations with the specific DTT parameters, to corroborate the validity of the reduced models in the particular case of DTT. In this paper, we summarise the first results of this activity, which extends the preliminary predictions reported in 1. The integrated modelling of DTT has been carried out with the JINTRAC suite [6] and covers the region inside the separatrix, whilst the values of temperature and density at the separatrix are taken consistently with the scrape-off layer simulations described in 1. The pedestal has been determined with the EPED1 model [7] implemented in the Europed code [8], and core-edge coupling has been taken into account on an iterative basis. The pedestal density has been set to achieve a volume averaged density ~ 0.43 nGW (Greenwald limit). The region in-side the top of the pedestal has been modelled using the QuaLiKiz [9] or the TGLF [10] turbulent transport models and NCLASS [11] for the neoclassical transport. The simulations pre-dict steady-state profiles of ion and electron temperature, density, rotation, current density, impurity (Ar, W) density, and calculate a self-consistent equilibrium starting from a fixed boundary taken from [12]. The heating has been modelled self-consistently using PENCIL[13] for NBI, PION[14] for ICRH and GRAY[15] for ECRH. SANCO [16] has been used to calculate impurity ionisation and recombination and radiation. The rotation has been predicted using a semi-empirical estimate of Prandtl and pinch numbers [17] due to numerical issues using the turbulent momentum transport from the quasi-linear models. Fig.1 shows profiles obtained for the SN full power H-mode scenario with 32 MW ECRH, 15 MW NBI and 3 MW ICRH using QuaLiKiz for turbulent transport, which is mainly driven by ion-scale ITG/TEM. The strong central ECRH peaks Te far above Ti in the central part. Ions are rather stiff and Ti stays below Te also in most of the outer region, in spite of a large amount of thermal exchange power from electrons to ions. The ne profile is moderately peaked. A peaked rotation profile with central value of 50 krad/s does not provide a significant ExB stabilisation of the ion heat transport. Global plasma parameters are ?N=1.6, ?E=0.28s, total DD neutron rate~1.4 1017 s-1 (30% thermal). Total radiation is 15 MW. Both Ar and W show peaked profiles, with hints of W central accumulation, however a better treatment of W neoclassical transport using NEO [18] is in plan to check this prediction. In this simulation MHD has not been included, but some considerations on MHD stability will be discussed. Similar simulations for the half-power DAY1 heating configuration yield very similar profiles and double confinement time, which is also an indication of the high stiffness. Simulations with TGLF are being finalised with the latest release of the model [19], and differences between the predictions by the two models will be assessed against gyrokinetic simulations using GENE [20].

Dedicated experiments to generate energetic D ions and D-(3) He fusion-born alpha particles were performed at the Joint European Torus (JET) with the ITER-like wall (ILW). Using the 3-ion D-(D-NBI)-(3) He radio frequency (RF) heating scenario, deuterium ions from neutral beam injection (NBI) were accelerated in the core of mixed D-(3) He plasmas to higher energies with ion cyclotron resonance frequency (ICRF) waves, in turn leading to a core-localized source of alpha particles. The fast-ion distribution of RF-accelerated D-NBI ions was controlled by varying the ICRF and NBI power (P-ICRF approximate to 4-6 MW, P-NBI approximate to 3-20 MW), resulting in rather high D-D neutron (approximate to 1x10(16) s(-1)) and D-(3) He alpha rates (approximate to 2x10(16) s(-1)) at moderate input heating power. Theory and TRANSP analysis shows that large populations of co-passing MeV-range D ions were generated using the D-(D-NBI)-(3) He 3-ion ICRF scenario. This important result is corroborated by several experimental observations, in particular gamma-ray measurements. The developed experimental scenario at JET provides unique conditions for probing several aspects of future burning plasmas, such as the contribution from MeV range ions to global confinement, but without introducing tritium. Dominant fast-ion core electron heating with T-i approximate to T-e and a rich variety of fast-ion driven Alfven eigenmodes (AEs) were observed in these D-(3) He plasmas. The observed AE activities do not have a detrimental effect on the thermal confinement and, in some cases, may be driven by the fusion born alpha particles. A strong continuous increase in neutron rate was observed during long-period sawteeth (>1 s), accompanied by the observation of reversed shear AEs, which implies that a non monotonic q profile was systematically developed in these plasmas, sustained by the large fast-ion populations generated by the 3-ion ICRF scenario.