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Under the AMC, a range of activities covering education, research and outreach are foreseen. One of them concerns education and the build-up of competence related to disarmament, and for that reason collaborative efforts have been ongoing during 2021 and 2022 to develop a PhD-level course in verification of nuclear test explosions, and to offer it during September-October 2022. The course has developed by Uppsala University and the Swedish Defence Reserach Agency (FOI) and corresponds to 7.5 credits. It is a cross-disciplinary course that spans over several disciplines. It introduces the participants to treaties and verification regimes governing nuclear weapons and it explains identification, calculation and analysis of signatures from nuclear weapon explosions. Furthermore, effort has been made to let the participants actively work with data collection, aggregation, analysis and with the interpretation and evaluation of data. The course includes also both a laboratory exercise on detection of radionuclides, and a project work in which the participants analyze a test explosion scenario and summarize their findings and conclusions in a manner very similar to how this is done in reality. This poster will describe the details of the course and its content. Since the course is planned to be offered just before this conference, we also hope to provide some information on its execution, as well as feedback from the participants.

A series of experiments was performed in the JET tokamak aiming to study the characteristics and eventual effects of beam injected ion populations further accelerated through 2nd harmonic ion cyclotron heating. It was found that the injection of these ions could affect sawtooth stability and that these populations excite toroidicity induced Alfvén eigenmodes (TAEs) in the core of the plasma. More interestingly, measurements of DD beam-plasma neutrons from the TOFOR spectrometer show that these modes caused local bump-on-tail distributions in energy. Numerical simulations performed with the CASTOR-K code found a strong interaction between the core-localized TAEs and ions with energies at which local minima in the energy distribution were measured.

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., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab3b31

Retrieving spatial distribution of plasma emissivity from line integrated measurements on tokamaks presents a challenging task due to ill-posedness of the tomography problem and limited number of the lines of sight. Modern methods of plasma tomography therefore implement a-priori information as well as constraints, in particular some form of penalisation of complexity. In this contribution, the current tomography methods under development (Tikhonov regularisation, Bayesian methods and neural networks) are briefly explained taking into account their potential for integration into the fusion reactor diagnostics. In particular, current development of the Minimum Fisher Regularisation method is exemplified with respect to real-time reconstruction capability, combination with spectral unfolding and other prospective tasks., For complete list of authors see http://dx.doi.org/10.1007/s10894-018-0178-x

Tritium accumulation in fusion reactor materials is considered a serious radiological issue, therefore a lot of effort has been concentrated on the development of radiometric techniques. A novel method, based on gradual dissolution, for the determination of the total tritium content and its depth profiles in metallic samples is demonstrated. This method allows for the measurement of tritium in metallic samples after their exposure to a hydrogen and tritium mixture, tritium containing plasma or after irradiation with neutrons resulting in tritium formation. In this method, successive layers of metal are removed using an appropriate etching agent in the controlled regime and the amount of evolved gases are measured by means of chromatography (gas composition and release rate) and a proportional gas flow detector (tritium). Results for the tritium profiles in neutron irradiated, plasma exposed and gas loaded beryllium are reported., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab3056

Developing successful tokamak operation scenarios, as well as confident extrapolation of present-day knowledge requires a rigorous understanding of plasma turbulence, which largely determines the quality of the confinement. In particular, accurate particle transport predictions are essential due to the strong dependence of fusion power or bootstrap current on the particle density details. Here, gyrokinetic turbulence simulations are performed with physics inputs taken from a JET power scan, for which a relatively weak degradation of energy confinement and a significant density peaking is obtained with increasing input power. This way physics parameters that lead to such increase in the density peaking shall be elucidated. While well-known candidates, such as the collisionality, previously found in other studies are also recovered in this study, it is furthermore found that edge E x B shearing may adopt a crucial role by enhancing the inward pinch. These results may indicate that a plasma with rotational shear could develop a stronger density peaking as compared to a non-rotating one, because its inward convection is increased compared to the outward diffusive particle flux as long as this rotation has a significant on E x B flow shear stabilization. The possibly significant implications for future devices, which will exhibit much less torque compared to present day experiments, are discussed., For complete list of authors see http://dx.doi.org/10.1088/1361-6587/ab31a4

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., For complete list of authors see http://dx.doi.org/10.1016/j.cpc.2019.05.007

For many years, machine learning tools have proved to be very powerful disruption predictors in tokamaks. On the other hand, the vast majority of the techniques deployed assume that the input data is independent and is sampled from exactly the same probability distribution for the training set, the test set and the final real time deployment. This hypothesis is certainly not verified in practice, since the experimental programmes evolve quite rapidly, resulting typically in ageing of the predictors and consequent suboptimal performance. This paper describes various adaptive training strategies that have been tested to maintain the performance of disruption predictors in non-stationary conditions. The proposed approaches have been implemented using new ensembles of classifiers, explicitly developed for the present application. The improvements in performance are unquestionable and, given the difficulties encountered so far in translating predictors from one device to another, the proposed adaptive methods from scratch can therefore be considered a useful option in the arsenal of alternatives envisaged for the next generation of devices, particularly at the very beginning of their operation., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab1ecc

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., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab318b

Alfven eigenmodes driven by energetic particles are routinely observed in tokamak plasmas. These modes consist of poloidal harmonics of shear Alfven waves coupled by inhomogeneity in the magnetic field. Further coupling is introduced by 3D inhomogeneities in the ion density during the assimilation of injected pellets. This additional coupling modifies the Alfven continuum and discrete eigenmode spectrum. The frequencies of Alfven eigenmodes drop dramatically when a pellet is injected in JET. From these observations, information about the changes in the ion density caused by a pellet can be inferred. To use Alfven eigenmodes for MHD spectroscopy of pellet injected plasmas, the 3D MILD codes Stellgap and AE3D were generalised to incorporate 3D density profiles. A model for the expansion of the ionised pellet plasmoid along a magnetic field line was derived from the fluid equations. Thereby, the time evolution of the Alfven eigenfrequency is reproduced. By comparing the numerical frequency drop of a toroidal Alfven eigenmode (TAE) to experimental observations, the initial ion density of a cigar-shaped ablation region of length 4cm is estimated to be n(*) = 6.8 x 10(22) m(-3) at the TAE location (r/a approximate to 0.75). The frequency sweeping of an Alfven eigenmode ends when the ion density homogenises poloidally. Modelling suggests that the time for poloidal homogenisation of the ion density at the TAE position is tau(h) = 18 +/- 4 ms for inboard pellet injection, and tau(h) = 26 +/- 2 ms for outboard pellet injection. By reproducing the frequency evolution of the elliptical Alfven eigemnode (EAE), the initial ion density at the EAE location (r/a approximate to 0.9) can be estimated to be n(*) = 4.8 x 10(22) m(-3). Poloidal homogenisation of the ion density takes 2.7 times longer at the EAE location than at the TAE location for both inboard and outboard pellet injection., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab382b

Deuterium pellets are injected into an initially pure hydrogen H-mode plasma in order to control the hydrogen: deuterium (H:D) isotope mixture. The pellets are deposited in the outer 20% of the minor radius, similar to that expected in ITER, creating transiently hollow electron density profiles. A H: D isotope mixture of approximately 45%:55% is obtained in the core with a pellet fuelling throughput of Phi(pel) = 0.045P(aux)/T-e,T-ped similar to previous pellet fuelling experiments in pure deuterium. Evolution of the H: D mix in the core is reproduced using a simple model, although deuterium transport could be higher at the beginning of the pellet train compared with the flat-top phase., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab3812

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., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab2ea9

Predictability of burning plasmas is a key issue for designing and building credible future fusion devices. In this context, an important effort of physics understanding and guidance is being carried out in parallel to JET experimental campaigns in H and D by performing analyses and modelling towards an improvement of the understanding of DT physics for the optimization of the JET-DT neutron yield and fusion born alpha particle physics. Extrapolations to JET-DT from recent experiments using the maximum power available have been performed including some of the most sophisticated codes and a broad selection of models. There is a general agreement that 11-15 MW of fusion power can be expected in DT for the hybrid and baseline scenarios. On the other hand, in high beta, torque and fast ion fraction conditions, isotope effects could be favourable leading to higher fusion yield. It is shown that alpha particles related physics, such as TAE destabilization or fusion power electron heating, could be studied in ITER relevant JET-DT plasmas., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab25b1

In plasmas heated with deuterium beams a deficit of the expected fusion neutron rate is an indicator of the deterioration of the fast-ion confinement, caused, for instance, by magnetohydrodynamic instabilities. The capability of predicting this deficit during the discharge relies on the availability of real-time estimates of the neutron rate from NBI codes which must be fast and accurate at the same time. Therefore, the recently developed real-time RABBIT code for neutral beam injection (NBI) simulations has been extended to output the distribution function and calculate the neutron emission. After the description of this newly installed diagnostics in RABBIT, benchmarks with NUBEAM, a massively used and validated Monte Carlo NBI solver, are discussed on ASDEX-Upgrade and JET cases. A first application for control-room intershot analysis on DIII-D is presented, and the results are compared on a large database with a slower NUBEAM analysis. Further application possibilities, e.g. for real-time control of Alfven eigenmodes, are outlined., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab1edd

This paper compares the gyrokinetic instabilities and transport in two representative JET pedestals, one (pulse 78697) from the JET configuration with a carbon wall (C) and another (pulse 92432) from after the installation of JET's ITER-like Wall (ILW). The discharges were selected for a comparison of JET-ILW and JET-C discharges with good confinement at high current (3 MA, corresponding also to low ρ*) and retain the distinguishing features of JET-C and JET-ILW, notably, decreased pedestal top temperature for JET-ILW. A comparison of the profiles and heating power reveals a stark qualitative difference between the discharges: the JET-ILW pulse (92432) requires twice the heating power, at a gas rate of 1.9 x 1022 e s-1, to sustain roughly half the temperature gradient of the JET-C pulse (78697), operated at zero gas rate. This points to heat transport as a central component of the dynamics limiting the JET-ILW pedestal and reinforces the following emerging JET-ILW pedestal transport paradigm, which is proposed for further examination by both theory and experiment. ILW conditions modify the density pedestal in ways that decrease the normalized pedestal density gradient a/Ln, often via an outward shift in relation to the temperature pedestal. This is attributable to some combination of direct metal wall effects and the need for increased fueling to mitigate tungsten contamination. The modification to the density profile increases η = Ln/LT, thereby producing more robust ion temperature gradient (ITG) and electron temperature gradient driven instability. The decreased pedestal gradients for JET-ILW (92432) also result in a strongly reduced E x B shear rate, further enhancing the ion scale turbulence. Collectively, these effects limit the pedestal temperature and demand more heating power to achieve good pedestal performance. Our simulations, consistent with basic theoretical arguments, find higher ITG turbulence, stronger stiffness, and higher pedestal transport in the, For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab25bd

Data on erosion and melting of beryllium upper limiter tiles, so-called dump plates (DP), are presented for all three campaigns in the JET tokamak with the ITER-like wall. High-resolution images of the upper wall of JET show clear signs of flash melting on the ridge of the roofshaped tiles. The melt layers move in the poloidal direction from the inboard to the outboard tile, ending on the last DP tile with an upward going waterfall-like melt structure. Melting was caused mainly by unmitigated plasma disruptions. During three ILW campaigns, around 15% of all 12376 plasma pulses were catalogued as disruptions. Thermocouple data from the upper dump plates tiles showed a reduction in energy delivered by disruptions with fewer extreme events in the third campaign, ILW-3, in comparison to ILW-1 and ILW-2. The total Be erosion assessed via precision weighing of tiles retrieved from JET during shutdowns indicated the increasing mass loss across campaigns of up to 0.6 g from a single tile. The mass of splashed melted Be on the upper walls was also estimated using the high-resolution images of wall components taken after each campaign. The results agree with the total material loss estimated by tile weighing (similar to 130 g). Morphological and structural analysis performed on Be melt layers revealed a multilayer structure of re-solidified material composed mainly of Be and BeO with some heavy metal impurities Ni, Fe, W. IBA analysis performed across the affected tile ridge in both poloidal and toroidal direction revealed a low D concentration, in the range 1-4 x 1017 D atoms cm-2., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab2076

The effect of fast ions on turbulent particle transport, driven by ion temperature gradient (ITG)/trapped electron mode turbulence, is studied. Two neutral beam injection (NBI) heated JET discharges in different regimes are analyzed at the radial position rho(t) = 0.6, one of them an L-mode and the other one an H-mode discharge. Results obtained from the computationally efficient fluid model EDWM and the gyro-fluid model TGLF are compared to linear and nonlinear gyrokinetic GENE simulations as well as the experimentally obtained density peaking. In these models, the fast ions are treated as a dynamic species with a Maxwellian background distribution. The dependence of the zero particle flux density gradient (peaking factor) on fast ion density, temperature and corresponding gradients, is investigated. The simulations show that the inclusion of a fast ion species has a stabilizing influence on the ITG mode and reduces the peaking of the main ion and electron density profiles in the absence of sources. The models mostly reproduce the experimentally obtained density peaking for the L-mode discharge whereas the H-mode density peaking is significantly underpredicted, indicating the importance of the NBI particle source for the H-mode density profile., For complete list of authors see http://dx.doi.org/10.1088/1361-6587/ab1e65

The self-consistent core-pedestal prediction model of a combination of EPED1 type pedestal prediction and a simple stiff core transport model is able to predict Type I ELMy (edge localized mode) pedestals of a large JET-ILW (ITER-like wall) database at the similar accuracy as is obtained when the experimental global plasma beta is used as input. The neutral penetration model [R. J. Groebner et al., Phys. Plasmas 9, 2134 (2002)] with corrections that take into account variations due to gas fueling and plasma triangularity is able to predict the pedestal density with an average error of 15%. The prediction of the pedestal pressure in hydrogen plasma that has higher core heat diffusivity compared to a deuterium plasma with similar heating and fueling agrees with the experiment when the isotope effect on the stability, the increased diffusivity, and outward radial shift of the pedestal are included in the prediction. However, the neutral penetration model that successfully predicts the deuterium pedestal densities fails to predict the isotope effect on the pedestal density in hydrogen plasmas., For complete list of authors see http://dx.doi.org/10.1063/1.5096870

NBI-heated L-mode plasmas have been obtained in JET with the Be/W ITER-like wall (JET-ILW) in H and D, with matched profiles of the dimensionless plasma parameters, rho*, nu*, beta and q in the plasma core confinement region and same T-i/T-e and Z(eff). The achieved isotope identity indicates that the confinement scale invariance principle is satisfied in the core confinement region of these plasmas, where the dominant instabilities are Ion Temperature Gradient (ITG) modes. The dimensionless thermal energy confinement time, Omega(i) tau(E,th), and the scaled core plasma heat diffusivity, A chi(eff)/B-T, are identical in H and D within error bars, indicating lack of isotope mass dependence of the dimensionless L-mode thermal energy confinement time in JET-ILW. Predictive flux driven simulations with JETTO-TGLF of the H and D identity pair is in very good agreement with experiment for both isotopes: the stiff core heat transport, typical of JET-ILW NBI heated L-modes, overcomes the local gyro-Bohm scaling of gradient-driven TGLF, explaining the lack of isotope mass dependence in the confinement region of these plasmas. The effect of E x B shearing on the predicted heat and particle transport channels is found to be negligible for these low beta and low momentum input plasmas., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab1ccd

Geodesic acoustic modes (GAMs) may generate strong oscillations in the radial electric field and therefore are considered as a possible trigger mechanism for the L-H transition. This contribution focuses on the characterization of GAMs in JET plasmas when approaching the L-H transition aiming at understanding their possible role in triggering the transition. GAM and turbulence characteristics are measured at the plasma edge using Doppler backscattering for different plasma current and line-averaged densities. The radial location of the GAM often moves further inside when neutral beam injection is applied possibly as a response to changes in the turbulence drive. GAMs are found to have modest amplitude at the transition except for high density discharges where GAMs are stronger, suggesting that the GAM is not responsible for facilitating the transition as the L-H power threshold also increases with density in the high density branch of the L-H transition. Our results suggest that the GAM alone does not play a leading role for causing the L-H transition at JET., For complete list of authors see http://dx.doi.org/10.1088/1361-6587/ab1e73

For the first time, the JET operation in deuterium-tritium (D-T) plasma, which is scheduled to take place on JET in 2020, will be performed in the ITER mix of plasma-facing component materials. In view of the preparation of the DT campaign (DTE2), several aspects of the plasma operation require significant improvements, such as a real-time protection of the first wall. The risk of damaging the metallic PFCs caused by beryllium melting or cracking of tungsten owing to thermal fatigue required a new reliable D-T compatible active protection system. Therefore, the future development of the JET real time first wall protection is focused on the D-T campaign and the ITER relevant conditions which may cause failure of camera electronics within the Torus hall. In addition to the technological aspect, the intensive preparation of the diverse software tools and real time algorithms for hot spot detection as well as alarm handling strategy required for the wall protection is in progress. This contribution describes the improved design, implementation, and operation of the near infrared (NIR) imaging diagnostic system of the JET-ILW plasma experiment and its integration into the existing JET protection architecture. To provide the reliable wall protection during the DTE2, two more sensitive logarithmic NIR camera systems equipped with new optical relays to take images and cameras outside of the biological shield have been installed on JET-ILW and calibrated with an in-vessel calibration light source (ICLS). Additionally, post-pulse data visualization and advanced analysis of all types of imaging data is provided by the new software framework JUVIL (JET users video imaging library). The formation of hot spots is recognized as a significant threat due to rapid surface temperature rise. Because it could trigger the protection system to stop a pulse, it is important to identify the mechanisms and conditions responsible for the formation of such hot spots. To address this issue the, For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab1a79

Results from two different sets of JET experiments are presented. In experiments in which toroidicity-induced Alfven eigenmodes (TAEs) localized at different radial locations had the same frequencies and toroidal mode numbers, the occurrence of enhanced losses after the excitation of TAEs in the core of the plasma was observed. On the contrary, enhanced losses were not observed if the TAEs localized at different radial locations had different frequencies and toroidal mode numbers. Numerical modeling indicates that, in the first set of experiments, the enhanced losses were caused by a combined effect of the TAEs localized at different radial locations. The TAEs localized in the plasma core transported energetic ions from the core to outer regions of the plasma. Then, the TAEs localized in outer regions of the plasma interacted with these ions just transported by the core-localized TAEs causing a further radial displacement of the ions to the plasma edge. This process eventually ends up causing the loss of the resonant ions. In the second set of experiments, it was found that TAEs localized in the plasma core and in outer regions did not interact with the same ions and so no enhanced losses were measured. Sheared profiles of the safety factor combined with flat mass density profiles lead to larger differences on the frequencies of the TAEs localized at different radial locations, eventually avoiding loss of energetic ions through the described mechanism., For complete list of authors see http://dx.doi.org/10.1088/1361-6587/ab27fd

The role of the pedestal position on the pedestal performance has been investigated in AUG, JET-ILW and TCV. When the pedestal is peeling-ballooning (PB) limited, the three machines show a similar behaviour. The outward shift of the pedestal density relative to the pedestal temperature can lead to the outward shift of the pedestal pressure which, in turns, reduces the PB stability, degrades the pedestal confinement and reduces the pedestal width. Once the experimental density position is considered, the EPED model is able to correctly predict the pedestal height. An estimate of the impact of the density position on a ITER baseline scenario shows that the maximum reduction in the pedestal height is 10% while the reduction in the fusion power is between 10% and 40% depending on the assumptions for the core transport model used. In other plasmas, where the pedestal density is shifted even more outwards relative to the pedestal temperature, the pedestal does not seem PB limited and a different behaviour is observed. The outward shift of the density is still empirically correlated with the pedestal degradation but no change in the pressure position is observed and the PB model is not able to correctly predict the pedestal height. On the other hand, the outward shift of the density leads to a significant increase of eta(e) and eta(i) (where eta(e,i) is the ratio of density to temperature scale lengths, eta(e,i) = L-eta e,L-i/L-Te,L-i) which leads to the increase of the growth rate of microinstabilities (mainly ETG and ITG) by 50%. This suggests that, in these plasmas, the increase in the turbulent transport due to the outward shift of the density might play an important role in the decrease of the pedestal performance., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab1eb9

The erosion of tungsten (W), induced by the bombardment of plasma and impurity particles, determines the lifetime of plasma-facing components as well as impacting on plasma performance by the influx of W into the confined region. The screening of W by the divertor and the transport of W in the plasma determines largely the W content in the plasma core, but the W source strength itself has a vital impact on this process. The JET tokamak experiment provides access to a large set of W erosion-determining parameters and permits a detailed description of the W source in the divertor closest to the ITER one: (i) effective sputtering yields and fluxes as function of impact energy of intrinsic (Be, C) and extrinsic (Ne, N) impurities as well as hydrogenic isotopes (H, D) are determined and predictions for the tritium (T) isotope are made. This includes the quantification of intra- and inter-edge localised mode (ELM) contributions to the total W source in H-mode plasmas which vary owing to the complex flux compositions and energy distributions in the corresponding phases. The sputtering threshold behaviour and the spectroscopic composition analysis provides an insight in the dominating species and plasma phases causing W erosion. (ii) The interplay between the net and gross W erosion source is discussed considering (prompt) re-deposition, thus, the immediate return of W ions back to the surface due to their large Larmor radius, and surface roughness, thus, the difference between smooth bulk-W and rough W-coating components used in the JET divertor. Both effects impact on the balance equation of local W erosion and deposition. (iii) Post-mortem analysis reveals the net erosion/deposition pattern and the W migration paths over long periods of plasma operation identifying the net W transport to remote areas. This W transport is related to the divertor plasma regime, e.g. attached operation with high impact energies of impinging particles or detached operation, as well as to the, For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab2aef

Convolutional neural networks (CNNs) have found applications in many image processing tasks, such as feature extraction, image classification, and object recognition. It has also been shown that the inverse of CNNs, so-called deconvolutional neural networks, can be used for inverse problems such as plasma tomography. In essence, plasma tomography consists in reconstructing the 2D plasma profile on a poloidal cross-section of a fusion device, based on line-integrated measurements from multiple radiation detectors. Since the reconstruction process is computationally intensive, a deconvolutional neural network trained to produce the same results will yield a significant computational speedup, at the expense of a small error which can be assessed using different metrics. In this work, we discuss the design principles behind such networks, including the use of multiple layers, how they can be stacked, and how their dimensions can be tuned according to the number of detectors and the desired tomographic resolution for a given fusion device. We describe the application of such networks at JET and COMPASS, where at JET we use the bolometer system, and at COMPASS we use the soft X-ray diagnostic based on photodiode arrays., For complete list of authors see http://dx.doi.org/10.1088/1748-0221/14/09/C09011

EDGE2D-EIRENE (the 'code') simulations show that radial electric field, Er, in the near scrape-off layer (SOL) of tokamaks can have large variations leading to a strong local E x B shear greatly exceeding that in the core region. This was pointed out in simulations of JET plasmas with varying divertor geometry, where the magnetic configuration with larger predicted near SOL E-r was found to have lower H-mode power threshold, suggesting that turbulence suppression in the SOL by local E. x. B shear can be a player in the L-H transition physics (Delabie et al 2015 42nd EPS Conf. on Plasma Physics (Lisbon, Portugal, 22-26 June 2015) paper O3.113 (http://ocs.ciemat.es/EPS2015PAP/pdf/O3.113.pdf), Chankin et al 2017 Nucl. Mater. Energy 12 273). Further code modeling of JET plasmas by changing hydrogen isotopes (H-D-T) showed that the magnitude of the near SOL E-r is lower in H cases in which the H-mode threshold power is higher (Chankin et al 2017 Plasma Phys. Control. Fusion 59 045012). From the experiment it is also known that hydrogen plasmas have poorer particle and energy confinement than deuterium plasmas, consistent with the code simulation results showing larger particle diffusion coefficients at the plasma edge, including SOL, in hydrogen plasmas (Maggi et al 2018 Plasma Phys. Control. Fusion 60 014045). All these experimental observations and code results support the hypothesis that the near SOL E x B shear can have an impact on the plasma confinement. The present work analyzes neutral ionization patterns of JET plasmas with different hydrogen isotopes in L-mode cases with fixed input power and gas puffing rate, and its impact on target electron temperature, T-e, and SOL E-r. The possibility of a self-feeding mechanism for the increase in the SOL E-r via the interplay between poloidal E x B drift and target T-e is discussed. It is also shown that reducing anomalous turbulent transport coefficients, particle diffusion and electron and ion heat conductivities, lead, For complete list of authors see http://dx.doi.org/10.1088/1361-6587/ab1629

Recent achievements in studies of the effects of ICRF (Ion Cyclotron Range of Frequencies) power on the SOL (Scrape-Off Layer) and PWI (Plasma Wall Interactions) in ASDEX Upgrade (AUG), Alcator C-Mod, and JET-ILW are reviewed. Capabilities to diagnose and model the effect of DC biasing and associated impurity production at active antennas and on magnetic field connections to antennas are described. The experiments show that ICRF near-fields can lead not only to E x B convection, but also to modifications of the SOL density, which for Alcator C-Mod are limited to a narrow region near antenna. On the other hand, the SOL density distribution along with impurity sources can be tailored using local gas injection in AUG and JET-ILW with a positive effect on reduction of impurity sources. The technique of RF image current cancellation at antenna limiters was successfully applied in AUG using the 3-strap AUG antenna and extended to the 4-strap Alcator C-Mod field-aligned antenna. Multiple observations confirmed the reduction of the impact of ICRF on the SOL and on total impurity production when the ratio of the power of the central straps to the total antenna power is in the range 0.6 < P-cen / P-total < 0.8. Near-field calculations indicate that this fairly robust technique can be applied to the ITER ICRF antenna, enabling the mode of operation with reduced PWI. On the contrary, for the A2 antenna in JET-ILW the technique is hindered by RF sheaths excited at the antenna septum. Thus, in order to reduce the effect of ICRF power on PWI in a future fusion reactor, the antenna design has to be optimized along with design of plasmafacing components., For complete list of authors see http://dx.doi.org/10.1016/j.nme.2018.11.017

The recently developed Monte-Carlo code ERO2.0 is applied to the modelling of limited and diverted discharges at JET with the ITER-like wall (ILW). The global beryllium (Be) erosion and deposition is simulated and compared to experimental results from passive spectroscopy. For the limiter configuration, it is demonstrated that Be self-sputtering is an important contributor (at least 35%) to the Be erosion. Taking this contribution into account, the ERO2.0 modelling confirms previous evidence that high deuterium (D) surface concentrations of up to similar to 50% atomic fraction provide a reasonable estimate of Be erosion in plasma-wetted areas. For the divertor configuration, it is shown that drifts can have a high impact on the scrape-off layer plasma flows, which in turn affect global Be transport by entrainment and lead to increased migration into the inner divertor. The modelling of the effective erosion yield for different operational phases (ohmic, L- and H-mode) agrees with experimental values within a factor of two, and confirms that the effective erosion yield decreases with increasing heating power and confinement., For complete list of authors see http://dx.doi.org/10.1016/j.nme.2019.01.015

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., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab2d4f

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., For complete list of authors see http://dx.doi.org/10.1016/j.nme.2019.02.009

Fishbones are ubiquitous in high-performance JET plasmas and are typically considered to be unimportant for scenario design. However, during recent high-performance hybrid scenario experiments, sporadic and explosive fishbone oscillations with sawtooth reconnection were observed coinciding with reduced performance and a main chamber hotspot. Fast ion loss diagnostics showed fusion products ejected from the plasma by the fishbones. We present calculations of the perturbed motion of non-resonant fusion products in the presence of fishbones assuming a fixed linear mode structure and frequency. Using careful reconstruction of the equilibrium and measurements of the perturbation, we show that the measured fishbone spatial structure in these experiments can be well modelled as a linear MHD internal kink mode. Both drift and full-orbit calculations predict losses of fusion products at the same location of the observed hotspot, however the calculated energy content of those losses is negligible and cannot be contributing significantly. The fast ions responsible for the hotspot and the reason for their loss both remain unexplained., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/aaea1e

The erosion, transport and deposition of tungsten in the outer divertor of JET-ILW has been studied for an H-Mode discharge with low frequency ELMs. For this specific case with an inter-ELM electron temperature at the strike point of about 20 eV, tungsten sputtering between ELMs is almost exclusively due to beryllium impurity and self-sputtering. However, during ELMs tungsten sputtering due to deuterium becomes important and even dominates. The amount of simulated local deposition of tungsten relative to the amount of sputtered tungsten in between ELMs is very high and reaches values of 99% for an electron density of 5E13 cm(-3) at the strike point and electron temperatures between 10 and 30 eV. Smaller deposition values are simulated with reduced electron density. The direction of the B-field significantly influences the local deposition and leads to a reduction if the E x B drift directs towards the scrape-off-layer. Also, the thermal force can reduce the tungsten deposition, however, an ion temperature gradient of about 0.1 eV/mm or larger is needed for a significant effect. The tungsten deposition simulated during ELMs reaches values of about 98% assuming ELM parameters according to free-streaming model. The measured WI emission profiles in between and within ELMs have been reproduced by the simulation. The contribution to the overall net tungsten erosion during ELMs is about 5 times larger than the one in between ELMs for the studied case. However, this is due to the rather low electron temperature in between ELMs, which leads to deuterium impact energies below the sputtering threshold for tungsten., For complete list of authors see http://dx.doi.org/10.1016/j.nme.2019.01.004

A synthetic diagnostic has been developed for the JET lost alpha scintillator probe, based on the ASCOT fast ion orbit following code and the AFSI fusion source code. The synthetic diagnostic models the velocity space distribution of lost fusion products in the scintillator probe. Validation with experimental measurements is presented, where the synthetic diagnostic is shown to predict the gyroradius and pitch angle of lost DD protons and tritons. Additionally, the synthetic diagnostic reproduces relative differences in total loss rates in multiple phases of the discharge, which can be used as a basis for total loss rate predictions., For complete list of authors see http://dx.doi.org/10.1088/1748-0221/14/09/C09018

Post-disruption runaway electron (RE) beams in tokamaks with large current can cause deep melting of the vessel and are one of the major concerns for ITER operations. Consequently, a considerable effort is provided by the scientific community in order to test RE mitigation strategies. We present an overview of the results obtained at FTU and TCV controlling the current and position of RE beams to improve safety and repeatability of mitigation studies such as massive gas (MGI) and shattered pellet injections (SPI). We show that the proposed RE beam controller (REB-C) implemented at FTU and TCV is effective and that current reduction of the beam can be performed via the central solenoid reducing the energy of REs, providing an alternative/parallel mitigation strategy to MGI/SPI. Experimental results show that, meanwhile deuterium pellets injected on a fully formed RE beam are ablated but do not improve RE energy dissipation rate, heavy metals injected by a laser blow off system on low-density flat-top discharges with a high level of RE seeding seem to induce disruptions expelling REs. Instabilities during the RE beam plateau phase have shown to enhance losses of REs, expelled from the beam core. Then, with the aim of triggering instabilities to increase RE losses, an oscillating loop voltage has been tested on RE beam plateau phase at TCV revealing, for the first time, what seems to be a full conversion from runaway to ohmic current. We finally report progresses in the design of control strategies at JET in view of the incoming SPI mitigation experiments., For complete list of authors see http://dx.doi.org/10.1088/1361-6587/aaef53

The first divertor was installed in the JET machine between 1992 and 1994 and was operated with carbon tiles and then beryllium tiles in 1994-5. Post-mortem studies after these first experiments demonstrated that most of the impurities deposited in the divertor originate in the main chamber, and that asymmetric deposition patterns generally favouring the inner divertor region result from drift in the scrape-off layer. A new monolithic divertor structure was installed in 1996 which produced heavy deposition at shadowed areas in the inner divertor corner, which is where the majority of the tritium was trapped by co-deposition during the deuterium-tritium experiment in 1997. Different divertor geometries have been tested since such as the Gas-Box and High-Delta divertors; a principle objective has been to predict plasma behaviour, transport and tritium retention in ITER. Transport modelling experiments were carried out at the end of four campaigns by puffing C-13-labelled methane, and a range of diagnostics such as quartz-microbalance and rotating collectors have been installed to add time resolution to the post-mortem analyses. The study of material migration after D-D and D-T campaigns clearly revealed important consequences of fuel retention in the presence of carbon walls. They gave a strong impulse to make a fundamental change of wall materials. In 2010 the carbon divertor and wall tiles were removed and replaced with tiles with Be or W surfaces for the ITER-Like Wall Project., For complete list of authors see http://dx.doi.org/10.1016/j.fusengdes.2018.10.002

This paper uses neutron emission spectrometry (NES), with the spectrometer TOFOR, to estimate the deuterium-tritium (DT) equivalent fusion yields of deuterium-only (DD) pulses at the tokamak JET. A method for making DT predictions using parameters determined from the neutron energy spectrum is described and this method is compared to corresponding estimations done with the modeling codes JESTORR and TRANSP, as well as with results from the 1997 JET DT campaign. The method has been applied to a large number of JET DD pulses conducted after the installation of the ITER-like wall (ILW), and the results have been used to assess the prospects for the upcoming DT campaign (DTE2). DT predictions made by NES produces similar estimations as JESTORR and TRANSP, and the fusion power from studied DT pulses fall in line with the estimated power of similar DD discharges. Extrapolating the recent JET pulses to estimate what the fusion yields would be in DTE2, results indicate that reaching over 10 MW is achievable and that the highest performing ILW pulses to date could approach the desired fusion power of 15 MW.

For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50%/50% D–T mixtures since 1997 and the first ever D–T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D–T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D–T preparation. This intense preparation includes the review of the physics basis for the D–T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D–T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the three-ions scheme), new diagnostics (neutron camera and spectrometer, active Alfvèn eigenmode antennas, neutral gauges, radiation hard imaging systems...) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D–T campaign provides an incomparable source of information and a basis for the future D–T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/ab2276

A high resolution neutron spectrometer (HRNS) system has been designed as a neutron diagnostic tool for ITER. The HRNS is dedicated to measurements of time resolved neutron energy spectra for both deuterium and deuterium-tritium (DT) plasmas. The main function of the HRNS is to determine the fuel ion ratio n(t)/n(d) in the plasma core with 20% uncertainty and a time resolution of 100ms for a range of ITER operating scenarios from 0.5 MW to 500 MW in fusion power. Moreover, neutron spectroscopy measurements should also be possible in the initial deuterium phase of ITER experiments. A supplementary function of the HRNS is to provide information on the fuel ion temperature. Furthermore, the HRNS can be used as an additional line-of-sight (LOS) for the radial neutron camera. To meet these requirements, a set of four spectrometers positioned after each other along a single LOS has been designed. The detector techniques employed include a thin foil proton recoil spectrometer (TPR), a neutron diamond detector (NDD), a back-scattering time-of-flight system (bToF) and a forward timeof-flight system (fToF). The TPR system, positioned closest to the plasma, provides data at high fusion powers. For plasma conditions producing intermediate fusion power two neutron spectrometers are installed: NDD and bToF. The NDD is installed as the second instrument along the HRNS LOS after the TPR. The fToF spectrometer is dedicated for low tritium densities and pure deuterium operation. The paper summarizes the current state of the art of neutron spectroscopy useful in plasma diagnostics and the possibility of installing a dedicated HRNS for ITER in the designated diagnostic port. We conclude that the proposed HRNS system can fulfil the ITER measurement requirements over a broad range of plasma operational scenarios, including full power DT, start-up, ramp-down and pure D operations.