Assessing the performance of tungsten plasma-facing components under disruption-relevant heat loads re-quires objective and reproducible methods to quantify surface damage in post-exposure microscopy images. In this work, we develop an automated framework for detecting and quantifying damage in scanning electron microscopy images of tungsten samples exposed at the OLMAT facility at CIEMAT, where leading-edge heat loads of up to 1200 MW/m2 are reproduced.The methodology follows three phases. A classical computer vision pipeline (TungstenAnalyzer) was first constructed, incorporating a dynamic topographic edge scanner, a multichannel Watershed engine, and a Frangi-Blackhat crack detector with dual hysteresis thresholding. A genetic multi objective optimisation algorithm (NSGA-II via Optuna) was then applied raising IoU over 261 generations. Recognising the intrinsic ceiling of handcrafted features, the third phase introduced a U-Net with a ResNet34 encoder pre-trained on ImageNet, augmented with scSE attention gates, deep supervision, and a crack-aware combined loss (BCE + Dice + Focal). Despite being trained on only 99 annotated SEM images, the model achieved an IoU of 0.839 for crater-like damage and 0.818 for cracks, corresponding to F1 scores of 0.912 and 0.688, respectively. The resulting multiscale inference workflow extracts physically meaningful quantities, including damaged area, crater severity and crack density, using automatic scale calibration from the SEM images.The initial analysis performed shows, as expected, that damage is systematically more severe at sample corners and leading edges than on equivalent flat regions, consistent with enhanced thermal loading due to local geometry. The complete workflow is integrated into a graphical user interface, providing a reproducible tool for routine quantification of high-heat-flux damage in tungsten plasma-facing materials for future works.

This study reports in-situ laser absorptance values at 1070 nm for solid tungsten samples, as well as liquid tin and lithium targets based on the tungsten capillary porous system technology. The effective absorptance has been estimated by matching transient thermal simulations based on the finite-element method to experimental temperature measurements at the OLMAT (Optimization of Liquid Metal Advanced Targets) facility. The targets have been characterized post-mortem by cross-sectional SEM-EDS, and the resulting microstructural and compositional data has been used to develop the numerical models. The resulting effective absorptance values have been obtained for different laser power levels and pulse lengths. In most cases, the effective absorptance decreases with increasing laser power and temperature, revealing significant discrepancies compared to reference values or pre-/post-exposure measurements, underlining the influence and complexities of the in-situ estimates. The results also indicate that the laser coupling coefficient is not constant under the tested conditions, which may be attributed to physical factors, such as surface state, wetting quality, and oxidation, as well as inherent modelling assumptions. Furthermore, we report the first dedicated effective absorptance values for liquid lithium, highlighting the critical role of the formation of passive oxide films. Quantifying this absorptance is essential to ensure that laboratory-scale energy deposition accurately reflects the thermal impact of real off-normal events. These results offer a consistent dataset for benchmarking the thermal response of liquid metal targets under laser-simulated fusion-relevant heat loads.

The armor for future nuclear fusion reactors is one of the main areas of research due to the harsh conditions it will undergo. Thermal fatigue is one of the most serious damage, as it will cause any material to fail even if the heat and particle loads during the reactor operation are always maintained low. In this work we have compared the actual tungsten armor for ITER tokamak with a new advanced tungsten material: tungsten reinforced by tungsten fibers (Wf/W). ITER-like W has shown small intergranular cracking at heat loads similar to the ones found in other similar devices: heat flux factor of FHF = 5.2 ± 1.6 MW/m2s0.5. But at much lower number of pulses: 641 versus 105. H embrittlement by the high-energy ions of OLMAT has been postulated as one of the main reason of this relatively prompt cracking appearance. Opposed to this, the type of Wf/W studied here, Porous Matrix (PM-Wf/W), has shown no damage at the same conditions and up to 950 pulses. These results show the capabilities of OLMAT for fatigue studies in conditions relevant to a future nuclear fusion reactor. New upgrades of OLMAT will partially solve the issues found in this first phase. Future work to continue with fatigue studies will be addressed.

Recently, nuclear fusion-related experiments and dedicated facilities worldwide relying on the usage of liquid metals are experiencing a renewed interest. This is the case of the International Fusion Materials Irradiation Facility – DEMO-Oriented Neutron Source (IFMIF-DONES), a radioactive facility which will use several cubic meters of pure liquid lithium as a target for deuterons to produce high neutron fluxes having an energy spectrum as effective as the expected fusion reactors. The primary mission of IFMIF-DONES will be to irradiate selected candidate structural materials for building a comprehensive database on fusion material properties. Generally, alkali metals, such as lithium, exhibit safety issues related to their chemical reactive nature and material compatibility. Due to the wide uncertainty regarding the lithium behavior confirmed by an extensive literature review, a dedicated experimental facility has been conceived to primarily study lithium ignition conditions in support of the licensing process of IFMIF-DONES. The ultimate purpose of the LiFIRE facility will be to demonstrate by means of experimental evidences the set of requirements on lithium fire protection to be applied to the final engineering design of IFMIF-DONES. This work describes the final design, construction and experimental campaign of the LiFIRE facility currently under development at CIEMAT's premises.

We provide an overview of activities carried out at the TJ-II stellarator for improving our understanding of- and developing plasma physics models for particle density profiles in stellarators. Namely, we report on recent progress in turbulent particle transport simulation, validation of pellet deposition models, density profile shaping for performance control and new experimental techniques for edge turbulence and plasma-neutral interaction.

The OLMAT (Optimization of Liquid Metal Advanced Targets) facility has recently undergone the commissioning and start-up phases. Solid Titanium -Zirconium-Molybdenum (TZM) alloy and liquid tin (Sn) metallic targets were exposed to a hydrogen neutral beam injector (NBI) particle flux with power densities up to 58 +/- 14 MW/ m2, pulse duration up to 150 ms, and repetition rates up to 2 pulses/minute. These beam parameters are well above the estimates based on the typical performance of this NBI system when used for heating plasmas in the TJ-II stellarator. The parameters of the plasma generated through the interaction of the fast (32.5 keV) neutrals and ions and the solid were characterized by spectroscopic methods while surface temperature and total absorbed power were followed using pyrometry, infrared (IR) thermography, and calorimetry, respectively. Targets were visually monitored during the exposure and microscopically analyzed ex-situ. Electrical isolation of the target permitted recording the floating voltage during irradiation as well as for active biasing tests. In this work, a description of the facility, its operating parameters, and firsts results are provided and assessed as a new High Heat Flux (HHF) Facility for testing solid and liquid metal divertor targets under reactor-relevant heat load conditions.

The operation of the Optimization of Liquid Metal Advanced Targets (OLMAT) facility began in April 2021 with the scientific objective of exposing liquid-metal plasma facing components (PFCs) to the particle and power fluxes provided by one of the hydrogen neutral beam injectors of the TJ-II stellarator. The system can deliver heat fluxes from 5 to 58 MW m −2 of high energy hydrogen neutral particles (≤ 33 keV) with fluxes up to 10 22 m 2 s −1 (containing an ion fraction ≤ 33% in some instances), pulsed operation of 30–150 ms duration and repetition rates up to 2 min −1 . These characteristics enable OLMAT as a high heat flux (HHF) facility for PFC evaluation in terms of power exhaust capabilities, thermal fatigue and resilience to material damage. Additionally, the facility is equipped with a wide range of diagnostics that includes tools for analyzing the thermal response of the targets as well as for monitoring atomic/plasma physics phenomena. These include spectroscopy, pyrometry, electrical probing and visualization (fast and IR cameras) units. Such particularities make OLMAT a unique installation that can combine pure technological PFC research with the investigation of physical phenomena such as vapor shielding, thermal sputtering, the formation/characterization of plasma plumes with significant content of evaporated metal and the detection of impurities in front of the studied targets. Additionally, a myriad of surface characterization techniques as SEM/EDX for material characterization of the exposed PFC prototypes are available at CIEMAT. In this article, first we provide an overview of the current facility upgrade in which a high-power CW laser, that can be operated in continuous and pulsed modes (0.2–10 ms), dump and electrical (single Langmuir) probe embedded on the target surface have been installed. This laser operation will allow simulating more relevant heat loading scenarios such as nominal steady-state divertor heat fluxes (10–20 MW m −2 in continuous mode) and transients including ELM loading and disruption-like events (ms time scales and power densities up to GW m −2 range). The work later focuses on the more recent experimentation (2022 fall campaign) where a 3D printed Tungsten (W) Capillary Porous System (CPS) target, with approximated 30 μm pore size and a 37% porosity and filled with liquid tin. This porous surface was a mock-up of the PFC investigated in the ASDEX Upgrade divertor manipulator. The target composed with this element was eventually exposed to a sequence of shots with the maximum heat flux that OLMAT provides (58 ± 14 MWm −2 ). Key questions as resilience to dry-out and particle ejection of the liquid metal layer, its refilling, the induced damage/modification of the porous W matrix and the global performance of the component are addressed, attempting to shed light on the issues encountered with the PFC at tokamak scale testing.

Four different tin-wetted, tungsten CPS (Capillary Porous System) targets where exposed to NBI pulses in the OLMAT High Heat flux (HHF) facility. They include two flexible ones placed on a TZM support (W meshes and W felt) and two compact ones (sintered W disk and 3D printed W). A comparative study was performed using a fastframe imaging camera and an infrared pyrometer. Surface temperature increase and homogeneity, particle ejection, CPS damage and overall behaviour were studied for each case. Sn drop/accumulation at the lower part of the targets was observed for all cases when they are heated up to around 400 C except for the 3D printed W target that has its own Sn deposit. The 3D printed W target presented the best results in all aspects, withstanding heat pulses up to 58 MW/m2 in 100 ms without any damage or particle ejection. On the other hand, the W mesh targets displayed damage at 20 MW/m2 due to a bad thermal contact with the deposit, while the sintered W disk developed a crack during a series of 15 MW/m2 NBI pulses. As might be expected, a reduced increase of temperature during pulses is observed for the two compact W targets. The results and their relevance for the design of a Sn wetted W CPS for application as a DEMO-divertor material are discussed in the present work.

In the DEMO-Oriented NEutron Source (DONES) facility, it is expected to use a considerable amount of lithium at an average temperature higher than 300 °C. It is well known that lithium can produce exothermic reactions with several media, generating corrosive and toxic (upon burning) reaction products and with the potential risk of mobilizing radionuclides in little amounts. As a result, lithium represents a hazardous material which must be safely handled under normal operation and, especially, potential accident conditions which involve lithium spills and radioactive releases.Although the burning behavior of lithium has been studied for several decades, there is still a wide dispersion of data regarding its actual ignition temperature. Generally, measurements of the spontaneous or induced ignition temperature of a leakage of molten lithium, pool/spray type, depend on several factors such as metal purity, gas atmosphere composition and relative humidity, sample size or geometrical aspects, and even different treatments, apparatus, and techniques used.A dedicated experimental facility, called “LiFIRE”, is currently under development at the National Fusion Laboratory of CIEMAT with the purpose to quantitatively support the DONES safety analysis on the fire risk in case of lithium leakages, based on the Defense-in-Depth principle, i.e. prevention, detection and mitigation.

OLMAT (Optimization of Liquid Metal Advanced Targets) is a new High Heat Flux (HHF) Facility for testing solid and liquid metallic targets at DEMO-relevant power densities that has been installed, commissioned and first operated at the National Fusion Laboratory in Madrid in 2021 [1] . A Neutral Beam Injector (NBI) is used as a high-power source providing 100 ms pulses at maximum repetition rate of 1p/30s and delivering power densities at the target (Mo, W, liquid Sn) above 50 MW/m 2 [2] . A calibrated survey spectrometer covering the UV-visible-near IR ranges provided information about the impinging hydrogenic species, plasma composition and its microscopic parameters, while a Balmer Ha (656.3 nm) monitor, a 16-Channel PMT array and a fast-frame imaging camera were used for the characterization of particle recycling and the plume’s spatial structure. Moreover, the floating voltage and saturation current developed at the target was monitored on a shot-to-shot basis. In addition to the above, nitrogen injection was performed through a dedicated piezoelectric valve for active spectroscopy. Plasma temperatures between 1 and 2 eV were deduced from the ratios of H Balmer lines, in line with the recorded values of floating potential, but higher, inconsistent values were obtained from the ratio of molecular N 2 and N 2 + bands. The facility and the plasmas therein generated will be described.

TJ-II stellarator results on modelling and validation of plasma flow asymmetries due to on-surface potential variations, plasma fuelling physics, Alfvén eigenmodes (AEs) control and stability, the interplay between turbulence and neoclassical (NC) mechanisms and liquid metals are reported. Regarding the validation of the neoclassically predicted potential asymmetries, its impact on the radial electric field along the flux surface has been successfully validated against Doppler reflectometry measurements. Research on the physics and modelling of plasma core fuelling with pellets and tracer encapsulated solid pellet injection has shown that, although post-injection particle radial redistributions can be understood qualitatively from NC mechanisms, turbulence and fluctuations are strongly affected during the ablation process. Advanced analysis tools based on transfer entropy have shown that radial electric fields do not only affect the radial turbulence correlation length but are also capable of reducing the propagation of turbulence from the edge into the scrape-off layer. Direct experimental observation of long range correlated structures show that zonal flow structures are ubiquitous in the whole plasma cross-section in the TJ-II stellarator. Alfvénic activity control strategies using ECRH and ECCD as well as the relation between zonal structures and AEs are reported. Finally, the behaviour of liquid metals exposed to hot and cold plasmas in a capillary porous system container was investigated.

Babinov N. Bader A. Badziak J. Baek S. Bagryansky P. Bakharev N. Ball J. Bandyopadhyay I. Banerjee S. Banerjee S. Banon Navarro A. Bao J. Barbisan M. Barbui T. Bastiani S. Batani D. Batistoni P. Battaglia D. Baylor L. Becoulet M. Belokurov A. Belonohy E. Belova E. Bergmann A. Berk H. Beurskens M. Bhattacharjee A. Bhattacharyay R. Biel W. Bierwage A. Biewer T. Bilato R. Bin W. Bisai N. Bisson R. Bizarro J. Blanchard J. Boeglin W. Bolzonella T. Bombarda F. Bonanomi N. Bonnin X. Bonoli P. Boozer A. Borodin D. Borodkina I. Borthakur S. Bortolon A. Bosch H.-S. Bourdelle C. Breizman B. Bremer P.-T. Brezinsek S. Briguglio S. Brizard A. Brochard G. Bromberg L. Brookman M. Browning P. Brunetti D. Brunsell P. Bruzzone P. Bühler L. Bufferand H. Buller S. Buratti P. Burhenn R. Buzi L. Byggmästar J. Bykov I. Bykov V.

In recent years it has been well known that in JET, with the ITER-like wall, the performance of high-power H-mode plasmas depends strongly on the magnetic topology of the divertor. This is generally attributed to the effect of the magnetic field shaping on the neutral flux transport and pumping, which-in high density H-mode plasmas-determine the pedestal properties and the global confinement. In this work we have analysed the spatial distribution and the dynamic behaviour of the D-alpha-emission for different magnetic configurations. Experimental observations indicate that for certain configurations, the surface temperature and the D-alpha-emission anomalously increase on top of the inner divertor, which points to thermal outgassing there. This is also the region where most beryllium co-deposits accumulate and most deuterium becomes trapped. The overheating at this region far from the strike point (SP) is observed to happen in magnetic configurations with reduced distance between the divertor material surface and the separatrix (clearance). The neutral flux that appears at the upper inner divertor during a few milliseconds after the ELM-crash, is more than an order of magnitude larger than the gas puffing rate and dominates over all other regions. Finally, a preliminary study describes how this thermal fuel outgassing from the co-deposited layers could be used intentionally as a wall-conditioning in JET technique with plasmas that focus their particle and heat flux there. This could be used as a complementary wall isotope control technique and more specifically for tritium recovery from the upper inner divertor where most fuel-trapping beryllium co-deposits accumulate in JET ITER-like wall.

In a future fusion reactor like DEMOnstration reactor (DEMO) one of the main concerns is the handling of the power exhaust from the plasma, especially at the divertor. The expected power loads cannot easily be handled by traditional armor solutions based on solid materials like tungsten, especially when the effect of intense neutron bombardment is also considered. Interest in armor concepts based on liquid metals has been subsequently on the rise, as they prove to be more resilient against high, fast power loads and neutron bombardment. However, engineering solutions for those concepts are very complex, and need to be tested. For this purpose, Optimization of Liquid Metal Advanced Targets project (OLMAT) has been envisaged. The project will use the Neutral Beam Injection of the TJ-II stellarator to irradiate liquid metal targets with power densities (neutrals plus occasionally ions) relevant to DEMO steady state operation, in the range of 20 MW/m 2 . OLMAT design will allow a series of experiments that other divertor simulator devices cannot easily perform: in-situ measurements of hydrogen retention, redeposition, vapor shielding, material fatigue, dust and precipitates effects, etc. Moreover, a high-power fiber laser will be used to simulate Edge Localized Modes in a small area, or to simulate the strike point power deposition profile.

Transport modelling of Joint European Torus (JET) dimensionless collisionality scaling experiments in various operational scenarios is presented. Interpretative simulations at a fixed radial position are combined with predictive JETTO simulations of temperatures and densities, using the TGLF transport model. The model includes electromagnetic effects and collisions as well as E ⃗ × B ⃗ shear in Miller geometry. Focus is on particle transport and the role of the neutral beam injection (NBI) particle source for the density peaking. The experimental 3-point collisionality scans include L-mode, and H-mode (D and H and higher beta D plasma) plasmas in a total of 12 discharges. Experimental results presented in (Tala et al 2017 44th EPS Conf. ) indicate that for the H-mode scans, the NBI particle source plays an important role for the density peaking, whereas for the L-mode scan, the influence of the particle source is small. In general, both the interpretative and predictive transport simulations support the experimental conclusions on the role of the NBI particle source for the 12 JET discharges.

Fusion performance in tokamaks hinges critically on the efficacy of the edge transport barrier (ETB) in suppressing energy losses. The new concept of 'fingerprints' is introduced to identify the instabilities that cause transport losses in the ETBs of many of today's experiments, from among widely posited candidates. Analysis of the gyrokinetic-Maxwell equations and gyrokinetic simulations of experiments reveals that each mode type produces characteristic ratios of transport in the various channels: density, heat, and impurities. This, together with experimental observations of transport in some channel or of the relative size of the driving sources of channels, can identify or determine the dominant modes causing energy transport. In multiple H-mode cases with edge-localized modes that are examined, these fingerprints indicate that magnetohydrodynamic (MHD)-like modes are apparently not the dominant agent of energy transport; rather, this role is played by micro-tearing modes (MTMs) and electron temperature gradient (ETG) modes, and in addition, possibly by ion temperature gradient/ trapped electron modes (ITG/TEM) on JET (Joint European 'Torus). MHD-like modes may dominate the electron particle losses. Fluctuation frequency can also be an important means of identification, and is often closely related to the transport fingerprint. The analytical arguments unify and explain previously disparate experimental observations on multiple devices, including DIII-D, JET, and ASDEX-U. Detailed simulations of two DIII-D ETBs also demonstrate and corroborate this.

In this paper, we analyze heat transport in the JET tokamak using data from its high resolution ECE diagnostic and analyses based on the transfer entropy (TE). The analysis reveals that heat transport is not smooth and continuous, but is characterized by 'trapping regions' separated by `minor transport barriers'. Meat may 'jump over' these barriers and when the heating power is raised, this 'jumping' behavior becomes more prominent. To check that our results are relevant for global heat transport, we deduced an effective diffusion coefficient from the TE results. Both its value and overall radial variation are consistent with heat diffusivities reported in literature. The detailed radial structure of the effective diffusion coefficient was shown to be linked to the mentioned minor transport barriers.

Results from two different sets of JET experiments are presented.In experiments in which toroidicity-induced Alfvén 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.