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Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment., QC 20240926

An overview of the modeling approaches, validation methods and recent main results of analysis and modeling activities related to the plasma-surface interaction (PSI) in JET-ILW experiments, including the recent H/D/T campaigns, is presented in this paper. Code applications to JET experiments improve general erosion/migration/retention prediction capabilities as well as various physics extensions, for instance a treatment of dust particles transport and a detailed description of melting and splashing of PFC induced by transient events at JET. 2D plasma edge transport codes like the SOLPS-ITER code as well as PSI codes are key to realistic description of relevant physical processes in power and particle exhaust. Validation of the PSI and edge transport models across JET experiments considering various effects (isotope effects, first wall geometry, including detailed 3D shaping of plasma-facing components, self-sputtering, thermo-forces, physical and chemically assisted physical sputtering formation of W and Be hydrides) is very important for predictive simulations of W and Be erosion and migration in ITER as well as for increasing quantitative credibility of the models. JET also presents a perfect test-bed for the investigation and modeling of melt material dynamics and its splashing and droplet ejection mechanisms. We attribute the second group of processes rather to transient events as for the steady state and, thus, treat those as independent additions outside the interplay with the first group., QC 20240905

An overview of recent results obtained at the tokamak ASDEX Upgrade (AUG) is given. A work flow for predictive profile modelling of AUG discharges was established which is able to reproduce experimental H-mode plasma profiles based on engineering parameters only. In the plasma center, theoretical predictions on plasma current redistribution by a dynamo effect were confirmed experimentally. For core transport, the stabilizing effect of fast ion distributions on turbulent transport is shown to be important to explain the core isotope effect and improves the description of hollow low-Z impurity profiles. The L-H power threshold of hydrogen plasmas is not affected by small helium admixtures and it increases continuously from the deuterium to the hydrogen level when the hydrogen concentration is raised from 0 to 100%. One focus of recent campaigns was the search for a fusion relevant integrated plasma scenario without large edge localised modes (ELMs). Results from six different ELM-free confinement regimes are compared with respect to reactor relevance: ELM suppression by magnetic perturbation coils could be attributed to toroidally asymmetric turbulent fluctuations in the vicinity of the separatrix. Stable improved confinement mode plasma phases with a detached inner divertor were obtained using a feedback control of the plasma beta. The enhanced D- alpha H-mode regime was extended to higher heating power by feedback controlled radiative cooling with argon. The quasi-coherent exhaust regime was developed into an integrated scenario at high heating power and energy confinement, with a detached divertor and without large ELMs. Small ELMs close to the separatrix lead to peeling-ballooning stability and quasi continuous power exhaust. Helium beam density fluctuation measurements confirm that transport close to the separatrix is important to achieve the different ELM-free regimes. Based on separatrix plasma parameters and interchange-drift-Alfven turbulence, an analytic mod, QC 20220406

An analysis workflow has been developed to assess energy deposition and material damage for ITER vertical displacement events (VDEs) and major disruptions (MD). This paper describes the use of this workflow to assess the melt damage to be expected during unmitigated current quench (CQ) phases of VDEs and MDs at different points in the ITER research plan. The plasma scenarios are modeled using the DINA code with variations in plasma current I (p), disruption direction (upwards or downwards), Be impurity density n (Be), and diffusion coefficient chi. Magnetic field line tracing using SMITER calculates time-dependent, 3D maps of surface power density q (perpendicular to) on the Be-armored first wall panels (FWPs) throughout the CQ. MEMOS-U determines the temperature response, macroscopic melt motion, and final surface topology of each FWP. Effects of Be vapor shielding are included. Scenarios at the baseline combination of I (p) and toroidal field (15 MA/5.3 T) show the most extreme melt damage, with the assumed n (Be) having a strong impact on the disruption duration, peak q (perpendicular to) and total energy deposition to the first wall. The worst-cases are upward 15 MA VDEs and MDs at lower values of n (Be), with q (perpendicular to,max) = 307 MW m(-2) and maximum erosion losses of similar to 2 mm after timespans of similar to 400-500 ms. All scenarios at 5 MA avoided melt damage, and only one 7.5 MA scenario yields a notable erosion depth of 0.25 mm. These results imply that disruptions during 5 MA, and some 7.5 MA, operating scenarios will be acceptable during the pre-fusion power operation phases of ITER. Preliminary analysis shows that localized melt damage for the worst-case disruption should have a limited impact on subsequent stationary power handling capability., QC 20220110

In nuclear fusion reactors, plasmas are heated to very high temperatures of more than 100 million kelvin and, in so-called tokamaks, they are confined by magnetic fields in the shape of a torus. Light nuclei, such as deuterium and tritium, undergo a fusion reaction that releases energy, making fusion a promising option for a sustainable and clean energy source. Tokamak plasmas, however, are prone to disruptions as a result of a sudden collapse of the system terminating the fusion reactions. As disruptions lead to an abrupt loss of confinement, they can cause irreversible damage to present-day fusion devices and are expected to have a more devastating effect in future devices. Disruptions expected in the next-generation tokamak, ITER, for example, could cause electromagnetic forces larger than the weight of an Airbus A380. Furthermore, the thermal loads in such an event could exceed the melting threshold of the most resistant state-of-the-art materials by more than an order of magnitude. To prevent disruptions or at least mitigate their detrimental effects, empirical models obtained with artificial intelligence methods, of which an overview is given here, are commonly employed to predict their occurrence—and ideally give enough time to introduce counteracting measures., QC 20230908

The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained and controlled W-melting experiment has been achieved for the first time in WEST on a poloidal sharp leading edge of an actively cooled ITER-like plasma facing unit (PFU). A series of dedicated high power steady state plasma discharges were performed to reach the melting point of tungsten. The leading edge was exposed to a parallel heat flux of about 100 MW.m(-2) for up to 5 s providing a melt phase of about 2 s without noticeable impact of melting on plasma operation (radiated power and tungsten impurity content remained stable at constant input power) and no melt ejection were observed. The surface temperature of the MB was monitored by a high spatial resolution (0.1 mm/pixel) infrared camera viewing the melt zone from the top of the machine. The melting discharge was repeated three times resulting in about 6 s accumulated melting duration leading to material displacement from three similar pools. Cumulated on the overall sustained melting periods, this leads to excavation depth of about 230 mu m followed by a re-solidified tungsten bump of 200 mu m in the JxB direction., QC 20211122

The beryllium (Be) main chamber wall interaction during a 5 MA/1.8 T upward, unmitigated VDE scenario, first analysed in [J. Coburn et al., Phys. Scr. T171 (2020) 014076] for ITER, has been re-evaluated using the latest energy deposition analysis software. Updates to the DINA disruption model are summarized, including an improved numerical convergence for the OD power balance, limitations on the safety factor within the plasma core, and the choice to maintain a constant plasma + halo poloidal cross-section. Such updates result in a broad halo region and higher radiated power fractions compared to previous models. The new scenario lasts for similar to 75 ms and deposits similar to 29 MJ of energy, with the radial distribution of parallel heat flux q parallel to(r) resembling an exponential falloff with an effective lambda(q) = 75 -198 mm. A maximum halo width w(h) of 0.52 m at the outboard midplane is observed. SMITER field line tracing and energy deposition simulations calculate a q(perpendicular to,max) of similar to 83 MW/m(2) on the upper first wall panels (FWP). Heat transfer calculations with the MEMOS-U code show that the FWP surface temperature reaches similar to 1000 K, well below the Be melt threshold. Variations of this 5 MA scenario with Be impurity densities from 0 to 3.10(19) m(-3) also remain below the melt threshold despite differences in energy deposition and duration. These results are in contrast to the early study which predicted melt damage to the first wall [J. Coburn et al., Phys. Scr. T171 (2020) 014076], and emphasize the importance of accurate models for the halo width w(h) and the heat flux distribution q parallel to(r) within that halo width. The 2020 halo model in DINA has been compared with halo current experiments on COMPASS, JET, and Alcator C-Mod, and the preliminary results build confidence in the broad halo width predictions. Results for the 5 MA VDE are compared with those for a 15 MA equivalent, generated using the new DINA model., QC 20210916

The MEMOS-U code, a significantly upgraded version of MEMOS-3D, has been developed to address macroscopic metallic melt motion in large-deformation long-displacement regimes, where melts spill onto progressively colder solid surfaces, that are ubiquitous in contemporary tokamaks and expected to be realized in ITER. The modelling of plasma effects, appearing via the free-surface boundary conditions, is discussed along with the sensitivity to external input. The crucial roles of convective and thermionic cooling are exemplified by simulations of ELM-induced tungsten leading edge melting. Key melt characteristics, revealed by previous MEMOS-U modelling of grounded sample exposures, are confirmed in new simulations of the recent floating sample exposures in ASDEX-Upgrade., QC 20210903

The MEMOS-U physics model, addressing macroscopic melt motion in large deformation and long displacement regimes, and its numerical schemes are presented. Discussion is centred on the shallow water application to the metallic melts induced by hot magnetized plasmas, where phase transitions and electromagnetic responses are pivotal. The physics of boundary conditions with their underlying assumptions are analysed and the sensitivity to experimental input uncertainties is emphasized. The JET transient tungsten melting experiment (Coenen et al 2015 Nucl. Fusion 55 023010) is simulated to illustrate the MEMOS-U predictive power and to highlight key aspects of tokamak melt dynamics., QC 20210215

Studies of macroscopic melt motion induced by fast transient power loads and the ensuing damage to plasma-facing components are critical for ITER. Based on state-of-the-art experiments, heat transfer is argued to be strongly entangled with fluid dynamics. The physics model of the MEMOS-U code is introduced. Simulations are reported of recent tokamak experiments concerning deliberate transient melting of beryllium main chamber tiles (JET) and tungsten divertor components (ASDEX Upgrade, JET). MEMOS-U is demonstrated to capture the main physics responsible for melt dynamics and to reproduce the observed surface deformation. The intricate role of resolidification is elucidated., QC 20201113

The beryllium (Be) first wall energy deposition and melt damage profiles resulting from the current quench phase of an unmitigated, 5 MA/1.8 T upward vertical displacement event for ITER are investigated. Time dependent 2D magnetic flux profiles are calculated with the DINA code and used as input for the SMITER 3D field line tracing software. 3D maps of the wetted area and perpendicular heat flux q(perpendicular to) show that the majority of the energy deposition occurs on the upper first wall panels #8 and #9 SMITER simulations predict q(perpendicular to,peak) approximate to 190 MW m(-2) on the surfaces of upper FWPs #8 and #9 at the end of the similar to 450 ms current quench. The surface heat flux maps generated by SMITER are used as input in the MEMOS-U code, which models Be melt formation and dynamics. Simulations reveal peak surface temperatures of similar to 2200 K, inward surface damage of similar to 0.5 mm in depth, and average melt velocities of similar to 2 m s(-1). Although VDEs are in principle the easiest disruptive instability to avoid, the analysis demonstrates that any non-mitigated events or intentional VDEs taking place during low I-p, early operational phases of ITER for the purposes of estimating disruption forces, must be kept to a low number., QC 20200914

The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q(95) = 5.5, beta(N) <= 2.8) at low density. Higher installed electron cyclotron resonance heating power <= 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG. Stable high-density H-modes with P-sep/R <= 11 MW m(-1) with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently H-H98(y,H-2) <= 1.1. This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated. Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of T-i and E-r allow for inter ELM transport analysis confirming that E-r is dominated by the diamagnetic term even for fast ti, QC 20191106

Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (n(e,sep)/n(G) similar to 0.3), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened., QC 20190807

The first MEMOS 3D simulations of liquid metal motion on an inclined bulk tungsten sample transiently molten by edge-localized modes (ELMs) are reported. The exposures took place at the outer ASDEX-Upgrade divertor with the tungsten surface tangent intersecting the magnetic field at similar to 18 degrees. Simulations confirm that the observed poloidal melt motion is caused by the volumetric J x B force with J the bulk replacement current triggered by thermionic emission. The final erosion profile and total melt build up are reproduced by employing the escaping thermionic current dependence on the incident heat flux derived from dedicated particle-in-cell simulations. Modelling reveals that melt dynamics is governed by the volumetric Lorentz force, capillary flows due to thermal surface tension gradients and viscous deceleration. The effect of the evolving surface deformation, that locally alters the field-line inclination modifying the absorbed power flux and the escaping thermionic current, in the final surface morphology is demonstrated to be significant., QC 20190118

The bulk replacement current density triggered by surface charge loss owing to thermionic emission leads to a volumetric Lorentz force which has been observed to drive macroscopic melt layer motion in transient tungsten melting tokamak experiments in which components of different geometries (deliberate leading edges and sloped surfaces) have been exposed to edge localized mode (ELM) pulsed heat loads in high power H-mode discharges. A self-consistent approach is formulated for the replacement current which is based on the magnetostatic limit of the resistive thermoelectric magnetohydrodynamic description of the liquid metal and results in a well-defined boundary value problem for the whole conductor. A new module is incorporated into the incompressible fluid dynamics code MEMOS-3D, which numerically solves the finite difference representation of the problem. The phenomenological approach, employed thus far to describe the replacement current, is demonstrated to be accurate for the sloped geometry but inadequate for the leading edge. MEMOS-3D simulations of very recent ASDEX-Upgrade leading edge experiments with the rigorous as well as the simplified approach are reported. For these simulations, the self-consistent approach predicts a fivefold reduction of the displaced material volume, a sevenfold reduction of the maximum peak height of displaced material and a different eroded surface morphology in comparison with the previously applied simplified approach., QC 20181218

The first comparison of dust radius and surface temperature estimates, obtained from spectroscopic measurements of thermal radiation, with simulations of dust heating and vaporization by the MIGRAINe dust dynamics code is reported. The measurements were performed during controlled tungsten dust injection experiments in the cold and dense plasmas of Pilot-PSI, reproducing ITER divertor conditions. The comparison has allowed us to single out the dominating role of the work function contribution to the dust heating budget. However, in the plasmas of interest, dust was found to enter the strong vaporization regime, in which its temperature is practically insensitive to plasma properties and the various uncertainties in modeling. This makes the dust temperature a poor figure of merit for model validation purposes. On the other hand, simple numerical scalings obtained from orbital-motion-limited estimates were found to be remarkably robust and sufficient to understand the main physics at play in such cold and dense plasmas., QC 20181002

Melt motion simulations of recent ASDEX Upgrade experiments on transient-induced melting of a tungsten leading edge during ELMing H-mode are performed with the incompressible fluid dynamics code MEMOS 3D. The total current flowing through the sample was measured in these experiments providing an important constraint for the simulations since thermionic emission is considered to be responsible for the replacement current driving melt motion. To allow for a reliable comparison, the description of the space-charge limited regime of thermionic emission has been updated in the code. The effect of non-periodic aspects of the spatio-temporal heat flux in the temperature distribution and melt characteristics as well as the importance of current limitation are investigated. The results are compared with measurements of the total current and melt profile., QC 20171127

Remobilization is one of the most prominent unresolved fusion dust-relevant issues, strongly related to the lifetime of dust in plasma-wetted regions, the survivability of dust on hot plasma-facing surfaces and the formation of dust accumulation sites. A systematic cross-machine study has been initiated to investigate the remobilization of tungsten micron-size dust from tungsten surfaces implementing a newly developed technique based on controlled pre-adhesion by gas dynamics methods. It has been utilized in a number of devices and has provided new insights on remobilization under steady-state and transient conditions. The experiments are interpreted with contact mechanics theory and heat conduction models., QC 20180111

QC 20210129