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).

TOFOR is a time-of-flight (TOF) neutron spectrometer located at the Joint European Torus (JET) with a vertical sightline of the JET plasma. It consists of 5 start (denoted S1) and 32 stop (denoted S2) plastic scintillation detectors which can be used in coincidence to generate a TOF spectrum. Spectroscopic analysis of the neutron TOF spectra produced by the JET plasma is regularly performed to determine, e.g., the fuel ion ratio and the presence of fast ion species in the fusion plasma. TOFOR has been upgraded with a new digital data acquisition (DAQ) system, denoted TOFu, which consists of 10 waveform digitizers with a total of 40 channels, 37 of which are connected to the photomultiplier output of the different S1 and S2 detectors. This paper presents a technical overview of the TOFu system and describes the offline analysis capabilities of TOFu which were not available with the previous DAQ system. Two experimental JET discharges are studied and used to show that the signal-to-background ratio is improved by almost 200% for the 2.5 MeV neutron signal and almost 400% for the 14 MeV neutron signal using the new offline analysis capabilities.

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 (alpha) 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-ILW plasma 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.

The TOFOR time-of-flight (TOF) neutron spectrometer at the Joint European Torus (JET) is composed of 5 start (S1) and 32 stop (S2) scintillation detectors. Recently, the data acquisition system (DAQ) of TOFOR was upgraded to equip each of the 37 detectors with its own waveform digitizer to allow for correlated time and pulse height analysis of the acquired data. Due to varying cable lengths and different pulse processing pathways in the new DAQ system, the 160 (5 · 32) different TOF pairs of start-stop detectors must be time-aligned to enable the proper construction of a summed TOF spectrum. Given the time (energy) resolution required by the entire spectrometer system to measure different plasma neutron emission components, it is of importance to align the detector pairs to each other with sub-nanosecond precision. Previously, the alignment partially depended on using fusion neutron data from Ohmic heating phases of JET experimental pulses. The dependence on fusion neutron data in the time alignment process is, however, unsatisfactory as it involves data one would wish to include in an independent analysis for physics results. In this work, we describe a method of time-aligning the detector pairs by using gamma rays. Given the known geometry and response of TOFOR to gamma rays, the time alignment of the detector pairs is found by examining gamma events interacting in coincidence in both S1-S1 and S1-S2 detector combinations. Furthermore, a technique for separating neutron and gamma events in the different detector sets is presented. Finally, the time-aligned system is used to analyze neutron data from Ohmic phases for different plasma conditions and to estimate the Ohmic fuel ion temperature.

An important instrumental development work has been done in the last two decades in the field of neutron and gamma ray spectroscopic measurements of magnetic confinement plasmas. Starting from the present state of the art instrumentation installed at JET, this paper reviews the recent development that has been carried out within the EUROFUSION programme for the forthcoming high power JET D and DT campaign. This development was dedicated to the realization of new compact neutron and gamma-ray spectrometers which combine very high energy resolution (typically better than 5%) and MHz counting rate capabilities allowing for time resolution in the 10 ms time scale. One of the advantages offered by the compact dimensions of these spectrometers is to make possible their use in multiple sight-line camera configurations, such as for future burning plasma reactors (ITER and DEMO). New compact neutron spectrometers based on single crystal diamond detectors have been developed and installed at JET for measurements of the 14 MeV neutron spectrum. Measurements on a portable DT neutron generator have shown that neutron spectroscopy of the accelerated beam ions at unprecedented energy resolution (∼1% at 14 MeV) is possible, which opens up new opportunities for diagnosing DT plasmas. For what concerns gamma ray measurements, the JET gamma ray camera has been recently upgraded with new compact spectrometers based on a LaBr3 scintillator coupled to Silicon Photomultiplier with the dual aim to improve the spectroscopic and rate capabilities of the detectors. The upgrade camera system will reconstruct the spatial gamma ray emissivity from the plasma in the MeV energy range at MHz counting rates and energy resolution in the 2–4% range. This will allow physics studies of gamma rays produced by the interaction of fast ions with impurities in the plasma and bremsstrahlung emission from runaway electrons.

This paper presents results of JET ITER-like wall L-mode experiments in hydrogen and deuterium (D) plasmas, dedicated to the study of the isotope dependence of ion heat transport by determination of the ion critical gradient and stiffness by varying the ion cyclotron resonance heating power deposition. When no strong role of fast ions in the plasma core is expected, the main difference between the two isotope plasmas is determined by the plasma edge and the core behavior is consistent with a gyro-Bohm scaling. When the heating power (and the fast ion pressure) is increased, in addition to the difference in the edge region, also the plasma core shows substantial changes. The stabilization of ion heat transport by fast ions, clearly visible in D plasmas, appears to be weaker in H plasmas, resulting in a higher ion heat flux in H with apparent anti-gyro-Bohm mass scaling. The difference is found to be caused by the different fast ion pressure between H and D plasmas, related to the heating power settings and to the different fast ion slowing down time, and is completely accounted for in non-linear gyrokinetic simulations. The application of the TGLF quasi-linear model to this set of data is also discussed.

The present paper offers an overview of the potential of ion cyclotron resonance heating (ICRH) or radio frequency heating for the DEMO machine. It is found that various suitable heating schemes are available. Similar to ITER and in view of the limited bandwidth of about 10 MHz that can be achieved to ensure optimal functioning of the launcher, it is proposed to make core second harmonic tritium heating the key ion heating scheme, assisted by fundamental cyclotron heating He-3 in the early phase of the discharge; for the present design of DEMO-with a static magnetic field strength of B-o = 5.855 T-that places the T and 3He layers in the core for f = 60 MHz and suggests centering the bandwidth around that main operating frequency. In line with earlier studies for hot, dense plasmas in large-size magnetic confinement machines, it is shown that good single pass absorption is achieved but that the size as well as the operating density and temperature of the machine cause the electrons to absorb a non-negligible fraction of the power away from the core when core ion heating is aimed at. Current drive and alternative heating options are briefly discussed and a dedicated computation is done for the traveling wave antenna, proposed for DEMO in view of its compatibility with substantial antenna-plasma distances. The various tasks that ICRH can fulfill are briefly listed. Finally, the impact of transport and the sensitivity of the obtained results to changes in the machine parameters is commented on.

The need for predictive capabilities greater than 95% with very limited false alarms are demanding requirements for reliable disruption prediction systems in tokamaks such as JET or, in the near future, ITER.The prediction of an upcoming disruption must be provided sufficiently in advance in order to apply effective disruption avoidance or mitigation actions to prevent the machine from being damaged.In this paper, following the typical machine learning workflow, a generative topographic mapping (GTM) of the operational space of JET has been built using a set of disrupted and regularly terminated discharges.In order to build the predictive model, a suitable set of dimensionless, machine-independent, physics-based features have been synthesized, which make use of 1D plasma profile information, rather than simple zero-D time series.The use of such predicting features, together with the power of the GTM in fitting the model to the data, obtains, in an unsupervised way, a 2D map of the multi-dimensional parameter space of JET, where it is possible to identify a boundary separating the region free from disruption from the disruption region.In addition to helping in operational boundaries studies, the GTM map can also be used for disruption prediction exploiting the potential of the developed GTM toolbox to monitor the discharge dynamics.Following the trajectory of a discharge on the map throughout the different regions, an alarm is triggered depending on the disruption risk of these regions.The proposed approach to predict disruptions has been evaluated on a training and an independent test set and achieves very good performance with only one tardive detection and a limited number of false detections.The warning times are suitable for avoidance purposes and, more important, the detections are consistent with physical causes and mechanisms that destabilize the plasma leading to disruptions.

The First Mirror Test for ITER has been carried out in JET with mirrors exposed during: (i) the third ILW campaign (ILW-3, 2015-2016, 23.33 h plasma) and (ii) all three campaigns, i.e. ILW-1 to ILW-3: 2011-2016, 63,52 h in total. All mirrors from main chamber wall show no significant changes of the total reflectivity from the initial value and the diffuse reflectivity does not exceed 3% in the spectral range above 500 nm. The modified layer on surface has very small amount of impurities such as D, Be, C, N, O and Ni. All mirrors from the divertor (inner, outer, base under the bulk W tile) lost reflectivity by 20-80% due to the beryllium-rich deposition also containing D, C, N, O, Ni and W. In the inner divertor N reaches 5 x 10(17) cm(-2), W is up to 4.3 x 10(17) cm(-2), while the content of Ni is the greatest in the outer divertor: 3.8 x 10(17) cm(-2). Oxygen-18 used as the tracer in experiments at the end of ILW-3 has been detected at the level of 1.1 x 10(16) cm(-2). The thickness of deposited layer is in the range of 90 nm to 900 nm. The layer growth rate in the base (2.7 pm s(-1)) and inner divertor is proportional to the exposure time when a single campaign and all three are compared. In a few cases, on mirrors located at the cassette mouth, flaking of deposits and erosion occurred.

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.

This paper outlines an approach towards improved rigour in tokamak turbulence transport model validation within integrated modelling. Gaussian process regression (GPR) techniques were applied for profile fitting during the preparation of integrated modelling simulations allowing for rigourous sensitivity tests of prescribed initial and boundary conditions as both fit and derivative uncertainties are provided. This was demonstrated by a JETTO integrated modelling simulation of the JET ITER-like-wall H-mode baseline discharge #92436 with the QuaLiKiz quasilinear turbulent transport model, which is the subject of extrapolation towards a deuterium-tritium plasma. The simulation simultaneously evaluates the time evolution of heat, particle, and momentum fluxes over similar to 10 confinement times, with a simulation boundary condition at rho(tor) = 0.85. Routine inclusion of momentum transport prediction in multi-channel flux-driven transport modelling is not standard and is facilitated here by recent developments within the QuaLiKiz model. Excellent agreement was achieved between the fitted and simulated profiles for n(e), T-e, T-i, and Omega(tor) within 2 sigma, but the simulation underpredicts the mid-radius Ti and overpredicts the core n(e) and T-e profiles for this discharge. Despite this, it was shown that this approach is capable of deriving reasonable inputs, including derivative quantities, to tokamak models from experimental data. Furthermore, multiple figures-of-merit were defined to quantitatively assess the agreement of integrated modelling predictions to experimental data within the GPR profile fitting framework.

ERO is a 3D Monte-Carlo impurity transport and plasma-surface interaction code. In 2011 it was applied for the ITER first wall (FW) life time predictions [1] (critical blanket module BM11). After that the same code was significantly improved during its application to existing fusion-relevant plasma devices: the tokamak JET equipped with an ITER-like wall and linear plasma device PISCES-B. This has allowed testing the sputtering data for beryllium (Be) and showing that the "ERO-min" fit based on the large (50%) deuterium (D) surface content is well suitable for plasma-wetted areas (D plasma). The improved procedure for calculating of the effective sputtering yields for each location along the plasma-facing surface using the recently developed semi-analytical sheath approach was validated. The re-evaluation of the effective yields for BM11 following the similar revisit of the JET data has indicated significant increase of erosion and motivated the current re-visit of ERO simulations.

Significant progress in applying ICRF power to ASDEX Upgrade (AUG) through improved power coupling and reduced impurity production has been associated with progress in understanding and modelling. First, the coupling of the fast wave was improved using outer mid-plane gas injection. The local edge density increase in front of the antenna shifts the fast wave cut-off position closer to the antenna by ~2cm, a result confirmed with new density measurements in front of the antenna. The ICRF coupling increases by 120% (only 25% for top gas puffing). The increased density and resulting improved coupling. was also numerically modelled. Second, the new 3-strap antennas in AUG demonstrated clearly that a proper antenna design can successfully mitigate ICRF-specific tungsten (W) sputtering. The reduction of the W sputtering was achieved by minimizing the RF currents on the antenna surfaces exposed to the scrape-off-layer (SOL) plasma. The local RF currents, rectified DC currents and the W sputtering yield at the antenna side limiters experience a clear minimum close to a phasing between the central and the outer straps of 180° and a power balance ratio Pcen/Pout of ~ 2. At this optimum, the local source of sputtered W at the limiters is reduced by a factor between 1.5 and 6, depending on the location. The experiments and modelling confirm the hypothesis of sheath rectification as the source of the sputtered W. Furthermore, the new 3-ion ICRF heating scenario, which can produce very energetic particles, has been successfully reproduced in AUG. The progress in operation, in understanding and modelling is strongly supported by improved ICRF diagnostic coverage including density measurements directly in front of the antenna by reflectometry, advanced RF coupling characterization, measurements of antenna limiter currents, B-dot probes, Ion Cyclotron Emission (ICE) measurements and by dedicated tests on the experimental device IShTAR (Ion cyclotron Sheath Test ARrangement).

Plasma tomography consists of reconstructing a two-dimensional radiation profile of a poloidal cross section of a fusion device based on line-integrated measurements along several lines of sight. The reconstruction process is computationally intensive, and in practice, only a few reconstructions are usually computed per pulse. In this work, we trained a deep neural network based on a large collection of sample tomograms that have been produced at JET over several years. Once trained, the network is able to reproduce those results with high accuracy. More importantly, it can compute all the tomographic reconstructions for a given pulse in just a few seconds. This makes it possible to visualize several phenomena—such as plasma heating, disruptions, and impurity transport—over the course of the entire pulse.

This paper presents the first application to real JET data of the new equilibrium code NICE which enables the consistent resolution of the inverse equilibrium reconstruction problem in the framework of non-linear free-boundary equilibrium coupled to the Stokes model equation for polarimetry. The conducted numerical experiments enable first of all to validate NICE by comparing it to the well-established EFIT code on 4 selected high performance shots. Secondly the results indicate that the fit to polarimetry measurements clearly benefits from the use of Stokes vector measurements compared to the classical case of Faraday measurements, and that the reconstructed p' and ff' profiles are better constrained with smaller error bars and are closer to the profiles reconstructed by EFTM, the EFIT JET code using internal MSE constraints.

The Neutron Camera Upgrade (NCU) is a neutron flux monitor consisting of six lines of sight (LoSs) under installation on Mega Ampere Spherical Tokamak (MAST) Upgrade. The NCU is expected to contribute to the study of the confinement of fast ions and on the efficiency of non-inductive current drive in the presence of on-axis and off-axis neutral beam injection by measuring the neutron emissivity profile along the equatorial plane. This paper discusses the NCU main design criteria, the engineering and interfacing issues, and the solutions adopted. In addition, the results from the characterization and performance studies of the neutron detectors using standard γ-rays sources and a 252Cf source are discussed. The proposed design has a time resolution of 1 ms with a statistical uncertainty of less than 10% for all MAST Upgrade scenarios with a spatial resolution of 10 cm: higher spatial resolution is possible by moving the LoSs in-between plasma discharges. The energy resolution of the neutron detector is better than 10% for a light output of 0.8 MeVee, and the measured pulse shape discrimination is satisfactory.

Progress in understanding lower hybrid current drive (LHCD) at high density has been made through experiments and modeling, which is encouraging given the need for an efficient off-axis current profile control technique in burning plasma. By reducing the wall recycling of neutrals, the edge temperature is increased and the effect of parametric instability (PI) and collisional absorption (CA) is reduced, which is beneficial for increasing the current drive efficiency. Strong single pass absorption is preferred to prevent CA and high LH operating frequency is essential for wave propagation to the core region at high density, presumably to mitigate the effect of PI. The dimensionless parameter that characterizes LH wave accessibility and wave refraction for the experiments in this joint study is shown to bracket the region in parameter space where ITER LHCD experiments will operate in the steady state scenario phase. Further joint experiments and cross modeling are necessary to understand the LHCD physics in weak damping regimes which would increase confidence in predictions for ITER where the absorption is expected to be strong.

Several important ELM control techniques are in large part motivated by the empirically observed inverse relationship between average ELM energy loss and ELM frequency in a plasma. However, to ensure a reliable effect on the energy released by the ELMs, it is important that this relation is verified for individual ELM events. Therefore, in this work the relation between ELM energy loss (WELM) and waiting time (ΔtELM) is investigated for individual ELMs in a set of ITER-like wall plasmas in JET. A comparison is made with the results from a set of carbon-wall and nitrogen-seeded ITER-like wall JET plasmas. It is found that the correlation between W ELM and ΔtELM for individual ELMs varies from strongly positive to zero. Furthermore, the effect of the extended collapse phase often accompanying ELMs from unseeded JET ILW plasmas and referred to as the slow transport event (STE) is studied on the distribution of ELM durations, and on the correlation between W ELM and ΔtELM . A high correlation between W ELM and ΔtELM , comparable to CW plasmas is only found in nitrogen-seeded ILW plasmas. Finally, a regression analysis is performed using plasma engineering parameters as predictors for determining the region of the plasma operational space with a high correlation between W ELM and ΔtELM .

A strong effect of divertor configuration on the threshold power for the L-H transition (PLH) was observed in recent JET experiments in the new ITER-like Wall (ILW) [1], [2], [3]. Following a series of EDGE2D-EIRENE code simulations with Be impurity and drifts a possible mechanism for the PLH variation with the divertor geometry is proposed. Both experiment and code simulations show that in the configuration with lower neutral recycling near the outer strike point (OSP), electron temperature (Te) peaks near the OSP prior to the L-H transition, while in the configuration with higher OSP recycling Te peaks further out in the scrape-off layer (SOL) and the plasma stays in the L-mode at the same input power. Code results show large positive radial electric field (Er) in the near SOL under lower recycling conditions leading to a large E × B shear across the separatrix which may trigger earlier (at lower input power) edge turbulence suppression and lower PLH. Suppressed Te's at OSP in configurations with strike points on vertical targets (VT) were observed earlier and explained by a geometrical effect of neutral recycling near this particular position, whereas in configurations with strike points on horizontal targets (HT) the OSP appears to be more open for neutrals (see e.g. review paper [4]).