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).

JET’s frequency-modulated continuous wave (FMCW) reflectometers have been operating well with the current design since 2005, and density profiles have been automatically calculated intershot since then. However, the calculated profiles had long suffered from several shortcomings: poor agreement with other diagnostics, sometimes inappropriately moving radially by several centimeters, elevated levels of radial jitter, and persistent wriggles (strong unphysical oscillations). In this research, several techniques are applied to the reflectometry data analysis, and the shortcomings are significantly improved. Starting with improving the equilibrium reconstruction that estimates the background magnetic field, adding a ripple correction in the reconstructed magnetic field profile, and adding new inner-wall reflection positions estimated through ray-tracing, these changes not only improve the agreement of reconstructed profiles to other diagnostics but also solve density profile wriggles that were present during band transitions. Other smaller but also persistent wriggles were also suppressed by applying a localized correction to the measured beat frequency where persistent oscillations are present. Finally, the burst analysis method, as introduced by Varela et al. [Nucl. Fusion 46 S693 (2006)], has been implemented to extract the beat frequency from stacked spectrograms. Due to the strong suppression of spurious reflections, the radial jitter that sometimes would span several centimeters has been strongly reduced. The stacking of spectrograms has also been shown to be very useful for stacking recurring events, like small gas puff modulations, and extracting transport coefficients that would otherwise be below the noise level.

Polarimetric Thomson scattering (PTS) is a technique that allows for accurate measurements of electron temperature (Te) in very hot plasmas (Te > 10 keV, a condition expected to be regularly achieved in ITER). Under such conditions, the spectral region spanned by the TS spectrum is large and extends to low wavelengths, where the transmission of the collection optics decreases, available detectors are less efficient, and the high level of plasma background light perturbs the measurements. This work presents the recent developments in the design of a PTS system for ITER, along with the challenges posed by the complex machine design. The system performance is assessed for an updated geometry (with respect to previous publication), showing that, with a scattering angle θscat = 167°, the expected signal is strongly reduced. Potential alternatives are analyzed: (1) a system employing a different laser injection position, allowing for a more favorable scattering angle and (2) a recently proposed dual-polarization laser pulse technique. The latter is evaluated for the possible ITER geometry, again showing that a more favorable scattering angle is needed for a robust performance.

In any large device for research on magnetic confinement thermonuclear fusion, the window assemblies are essential to preserve the conditions for successful experiments and to guarantee adequate access for inspection and measurement. In ITER, the materials traditionally used for the windows will be exposed to an exceptionally harsh environment. Moreover, the systematic use of tritium as fuel would make any failure of the primary vacuum containment a particularly dangerous accident. It is therefore essential to understand the potential threats to the integrity of the windows assemblies and define a series of tests to ensure their properties and quality before installation. One specific hazard to the windows is the microwave radiation due to either heating schemes or specific diagnostics. A fraction of the stray radiation incident on the window leads to dielectric heating, which causes a thermal load. The potentially harmful consequences of such a thermal load on the window assemblies are: (i) excessively high temperature (risk of bonding melting) and (ii) excessive, heating rate (risk of cracking due to a high temperature gradient at the location of the bonding). In this paper the main causes of degradation, which could lead to failures under microwave loads, have been identified. A series of laboratory tests have been defined, to assess the quality of the materials and the assemblies, including the coatings for the absorption of the microwave radiation in the ducts leading to the windows. Complete testing procedures and an overview of the main facilities, where the assemblies and materials could be qualified, are also provided.

In tokamaks with tungsten-based plasma facing components, such as ITER, pollution of the plasma by heavy impurities is a major concern as it can lead to radiative breakdown. The radiation emitted by such impurities is mainly composed of x-rays in the [0.1; 100] keV range. A diagnostic allowing for the reconstruction of the impurity distribution is of high interest. The ITER requirements for the x-ray measurement system make it mandatory for the detector to provide spectral information. Due to the radiation environment during the ITER nuclear phase, advanced detectors exhibiting high resilience to neutrons and gamma rays, such as gas-filled detectors, are required. The use of Low Voltage Ionization Chambers (LVICs) for this purpose is investigated in this paper. Several anodes have been added to the detector in order to allow for spectral deconvolution. This article presents a conceptual study of the use of a multi-anode LVIC for energy resolved x-ray measurement on ITER. It covers the design of the multi-anode LVIC and its modeling, the method for spectral deconvolution, and its application to energy resolved x-ray tomography, as well as the computation of the electron temperature from the reconstructed local x-ray emissivity.

ITER First Plasma (FP) operation aims to produce a plasma with current higher than 100 kA for a duration longer than 100 ms. However, the low values of plasma density and temperature associated with the initiation phase make it difficult to diagnose the plasma accurately. It is therefore essential to develop models for the available diagnostics for ITER FP operation in order to determine the necessary measurement ranges for plasma initiation and use them as inputs for controller development and assessment within the plasma control system (PCS). The use of accurate diagnostic models also helps to optimally prepare for, and analyse, ITER FP operation. A model developed for the Hα main chamber visible spectroscopy for FP shows that it is possible to measure Hα emission soon after breakdown for plasma temperatures higher than 3 eV and average plasma densities higher than 1017 m−3. The diagnostic model for the interferometer shows that for the proposed ITER FP scenarios, accurate measurements are possible for average plasma densities higher than 5 × 1017 m−3. The results from modelling of the hard x-ray monitor demonstrate that it can provide signals useful for runaway electron detection and can measure current levels typically down to 3 kA. Thus, these measurements can thus be used effectively by the PCS for runaway electron (RE) detection during FP operation. Improved diagnosis of the plasma state deducing, for example, parameters such as the plasma size and position, is possible by combining the measurements from the different FP diagnostics (density interferometer polarimeter and Hα main chamber visible spectroscopy), which will be available for ITER FP operation.

Here we re-evaluate the stray radiation loads due to electron cyclotron resonance heating and diagnostic gyrotrons in ITER, as well as compute the stray radiation energy flux due to electron cyclotron emission. Microwave loads due to stray radiation were assessed in the past. We give here more comprehensive estimates, including spatial variations in the machine. We show that for the start-up phase in the First Plasma campaign the threshold of 100 kW/m2, which diagnostics have to be able to withstand, is not exceeded. For the later campaigns, the stray radiation energy flux during start-up and burnthrough will exceed 450 kW/m2. Stray radiation from electron cyclotron emission and from a diagnostic 60 GHz gyrotron can contribute to more than 140 kW/m2 combined.

A dispersion interferometer (DI) is known to be less sensitive to mechanical vibrations, which is one of the main sources of error for conventional interferometers. A simple optical configuration is also one of the advantages of the DI approach. Since the first application of a homodyne DI on a nuclear fusion plasma device in the 1990s, several interferometer techniques have improved the operation of the DI. Improvements in nonlinear crystals have also improved the performance and availability of the DI. Immunity to neutral gas density changes, which is needed for low-temperature plasma measurements, is also found. A remaining challenge is the suppression of the offset drifts, which can be significant. Recent studies confirmed that they are caused by ambient humidity changes.

Soft x-ray (SXR) tomography is a key diagnostic method for impurity transport study in tokamaks since it allows for local impurity density reconstruction. The International Thermonuclear Experimental Reactor (ITER) radiative environment in deuterium-deuterium and deuterium-tritium phases will limit the choices of SXR detector technologies, and gas detectors are one of the most promising solutions. In this paper, we, thus, investigate the SXR tomography possibilities on ITER using Low Voltage Ionization Chambers (LVICs). The study contains the development of a LVIC synthetic diagnostic and its application to estimate the LVIC tomographic capabilities in an ITER D-T scenario, including the influence of LVIC parameters and noise in the measurements.

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.

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 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.

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.

In ITER, the window assemblies will be subjected to the combination of loads due to stray radiation and neutron streaming. Therefore, adequate tests are needed to qualify the windows for the hostile environmental conditions. In ITER, two main systems will generate high power waves in the millimeter range: the Collective Thomson Scattering (CTS) system, with a total power of 1 MW at 60 GHz, and the Electron Cyclotron Resonant Heating (ECRH) system, with a maximum power of 20 MW at 170 GHz. In particular operating conditions, a significant power fraction can be deflected in other ports as stray radiation. High levels of stray radiation could damage diagnostic windows that are safety relevant components. The radiation absorbed by the fused silica window assemblies on ports #11 and #12 have been evaluated in detail. The possible effects of gamma and neutron flux on the dielectric permittivity and loss tangent, determining the absorption, have also been reviewed.

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.

The ITER Integrated Modelling & Analysis Suite (IMAS) is the software infrastructure being developed using expertise from across the research institutes within the ITER Members to meet the needs of the ITER Integrated Modelling Programme. It builds around a standardised representation of data described by a Data Dictionary that is both machine independent and extensible. Machine independence is important for allowing tools developed in IMAS to be tested on existing devices, whilst extensibility allows the Data Dictionary to grow and evolve over time as more and more Use Cases are addressed. In addition to providing all the scientific tools for the exploitation of ITER once operations start, IMAS also has a role to play during the construction phase by providing simulation data to support systems design, in particular for diagnostics, heating, fuelling and control systems. An IMAS workflow developed from the DINA and JINTRAC code suites and coupling the transport across the core, edge and SOL regions has now, for the first time, performed simulations of the 15 MA / 5.3 T Q=10 DT ITER baseline scenario from the X-point formation during ramp-up to the X-point to limiter transition in the ramp-down. Recently an IMAS database of ITER simulations has started to be developed to help manage the exchange of data with ITER collaborators and Domestic Agencies. The database is being populated by translating existing data and running new simulations. The codes used to initially populate the database are ASTRA, CORSICA, DINA, JINTRAC, and METIS. Additional data structures consistently describing other aspects of the scenarios are being added to the database upon request. Scenarios for all stages of the ITER Research Plan are represented in the database which is accessible to all ITER contributors through the IMAS Access Layer, either for visualisation or as input to IMAS-adapted workflows and codes.

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.

Numerical scan at constant β shows that core and scrape-off layer (SOL) radiations do not depend on the plasma current (Ip). Whereas the SOL radiation increases with seeding, the core radiation, however, does not continue to increase with seeding but rolls over at higher seeding rates in the simulations. The core plasma contamination by W ions is low, cW ≪ 10−4. When the seeding starts, an increase in radiation power leading to a reduction in Ploss = (Paux − Prad) is observed, influencing the plasma confinement. The power scan at constant Ip indicates that the core radiation, Pplate, PSOL (and even SOL radiation), saturates with seeding. In addition, strong dilution with increasing seeding (Zeff ≫ 3) and large W concentrations with increasing power are found. Comparing neon with nitrogen seeding, it is seen that neon leads to slightly larger total radiation than nitrogen. However, that is achieved with much higher plasma contamination (Zeff ∼ 4–5) and dilution in the case of Ne, and simultaneously the power crossing the separatrix is lower for Ne than for N, indicating better H-mode performance in N2-seeded discharges.