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Divertor tiles after Joint European Torus-ITER like wall (JET-ILW) campaigns and dust collected after JET-C and JET-ILW operation were examined by a set of complementary techniques (full combustion and radiography) to determine the total, specific and areal tritium activities, poloidal tritium distribution in the divertor and the presence of that isotope in individual dust particles. In the divertor tiles, the majority of tritium is detected in the surface region and, the areal activities in the ILW divertor are in the 0.5-12 kBq cm-2 range. The activity in the ILW dust is associated mainly with the presence of carbon particles being a legacy from the JET-C operation. The total tritium activities show significant differences between the JET operation with ILW and the earlier phase with the carbon wall (JET-C) indicating that tritium retention has been significantly decreased in the operation with ILW.

DEMO will have a limited diagnostic set for optimization of reactor performance, and limited diagnostic coverage due to challenging reactor conditions. This poses challenges for control, especially for detachment control which is planned to be performed with a limited set of spectroscopic lines-of-sight. To demonstrate mitigation of the coverage problem, we have implemented a sparse sensor placement technique on TCV for the case of spectroscopy. Experimental CIII 2D emissivity profiles from the MANTIS multispectral imaging system at TCV are used to create a synthetic spectroscopy system, with DEMO-relevant lines-of-sight. This synthetic diagnostic is validated through comparison with spectroscopic measurements in TCV. The sparse sensor placement algorithm takes as input the radiance calculated for a large set of lines-of-sight, resolved in time. It provides a method to reconstruct an emission profile along the divertor leg based on measurements from only a few lines-of-sight. We demonstrate that it is feasible to use a calibration based on either experimental data or simulated emissivities from SOLPS-ITER. As a result, for DEMO, the line-of-sight selection and a mathematical basis for reconstruction can be obtained before the reactor is turned on. The sparse sensor placement method can also be applied to estimate other physical quantities from emission measurements, if they have a one-to-one relationship with the shape of the emission profile along the divertor leg. As an example, the peak target current density obtained with Langmuir probes can be reconstructed using spectroscopy measurements. Lastly, the geometry of the lines-of-sight plays an important role in proper emission profile reconstruction. It should be tailor-made to the used magnetic configuration to measure the emission profile accurately, and the technique described here offers a route for performing this optimization.

The consequences of tungsten (W) damaging processes, such as cracking and melting, on divertor lifetime and plasma operation are high priority issues for ITER. A sustained melting experiment was conducted in WEST using a 2 mm deep groove geometry on the upstream mono-block (MB) to overexpose the sharp leading edge (LE) of the downstream MB. W-cracking has been evidenced for the first time with a very high spatial resolution infrared camera before tungsten melting was reached. These cracks develop when the monoblock temperature is about 2600 degrees C, thus higher than both ductile to brittle transition and softening threshold of tungsten, suggesting that these cracks are different from the ones observed in previous campaigns where brittle failure was involved, because of transient events on cold monoblock. Post-exposure analyses have been performed on the damaged monoblock, highlighting 12 main cracks on the LE, with a width varying from 33 mu m to 77 mu m, and an average spacing of 0.45 mm. Parallel heat flux about 90 MW/m2 has been derived from infrared temperature measurements, with a heat flux decay length on the target of 4 mm. The T-REX modelling code suggest here that with these thermal inputs, a crack can initiates due to thermal cycling without disruption, with a ductile failure, under 1 to 5 cycles for a tungsten DBTT varying from 400 degrees C to 500 degrees C., QC 20231215

In this dissertation, we investigate two proposed features aimed at improving the performance of the divertor exhaust system in tokamak nuclear fusion devices. Specifically, we assess the influence of the "inverse" plasma sheath and long-leg divertor configurations on the divertor plasma using numerical simulations. First, we investigate the "inverse" plasma sheath, which has been suggested to prevent the flow of ions to the wall and promote divertor detachment. We use the UEDGE code to simulate the physics of both the inverse and standard (Bohm) sheath regime at the divertor targets. Our results show little difference in the overall plasma state, but an increase in electron heat flux to the divertor targets. Additionally, we observe a bifurcation behavior related to plasma recombination effects, and present an analytical model of this behavior. Second, we evaluate the long outer divertor leg, which is designed to increase volumetric dissipation, enhance turbulence spreading, and extend the distance between the material surface and the fragile core plasma. Using the SOLPS4.3 code, we assess the transition to the detached divertor regime, with scans on plasma density, transport coefficients, and multiple impurity species. Our results show a significant contribution to the energy balance from cross-field transport to the side walls and considerable recycling of impurities along the long leg, enabling delocalization of the radiated heat flux. Overall, this dissertation provides important insights into the physics associated with these proposed divertor features. The results suggest that the inverse plasma sheath may not have a significant impact on divertor plasma detachment, while the long-leg divertor configuration has potential to improve divertor performance by reducing impurity radiation localization and enhancing energy dissipation through cross-field transport. These findings contribute to the ongoing efforts to develop more efficient and effective divertor exha

The main goal of the Divertor Tokamak Test facility (DTT) is to operate with a high value of power-exhaust-relevant parameter Psoz/R in plasma scenarios similar to those foreseen for the Demonstration Fusion Power Plant (DEMO) in terms of low collisionality and neutral opacity. For these unique characteristics, accurate modelling of the principal scenario is necessary for machine designing. In edge numerical codes, cross-field transport profiles have a high impact on modelling results. This work aims at providing a coherent set of transport parameters for DTT full-power (FP) single-null (SN) scenario edge modelling. To evaluate such parameters for DTT, a transport analysis on the current machine has been performed using SOLEDGE2D-EIRENE and SOLPS-ITER. The transport parameters to be used in the simulations of the DTT single-null scenario were selected using two complementary methods. The first is the modelling of JET and Alcator C-Mod (C-Mod) with SOLEDGE2D-EIRENE and SOLPS-ITER, validating transport parameters by comparing modelling results to experimental data from pulses which are considered DTT-relevant. JET pulses were selected with the highest auxiliary input power (from 29 to 33 MW), plasma current and toroidal field to better match DTT parameters; nitrogen and neon seeded pulses were selected to check possible seeding material dependencies. The considered C-Mod pulse better matches DTT plasma density and neutral opacity. Transport parameters are then scaled to DTT according to scaling laws. The second method derives the transport parameters by tuning their values inside the DTT separatrix to reproduce the pedestal profiles predicted by the EPED model via the Europed code and applied in DTT. The derived set of DTT transport parameters is consistent with the results obtained by modelling present machines, reproduces the expected heat flux decay length in detached conditions and, inside the separatrix, reproduces the predicted pedestal using transport parameter, QC 20230607

Analysis of the divertor edge localized mode (ELM) electron temperature at a uniquely high temporal resolution (10(-5) s) was reported at the JET tokamak (Guillemaut et al 2018 Nucl. Fusion 58 066006). By collecting divertor probe data obtained during many dozens of ELMs, the conditional-average (CAV) technique yields surprisingly low peak electron temperatures, far below the pedestal ones (70%-99% reduction!) which we, however, question. This result was interpreted through the collisional free-streaming kinetic model of ELMs, by a transfer of most of the electron energy to ions, implying a high tungsten sputtering for unmitigated ELMs in future fusion devices like ITER. Recently, direct microsecond temperature measurements on the COMPASS tokamak, however, showed that the electron temperature peak of ELM filaments measured in the divertor is reduced by less than a third with respect to the pedestal one. This was further confirmed by a dedicated 1D particle-in-cell (PIC) simulation and tends to prove that the pedestal electrons can transfer only their parallel energy to ions (due to low collisionality), thus less than a third, as is predicted by the collisionless free-streaming model. This finding strongly contradicts the JET observations. We have therefore compared the CAV to the direct (microsecond) ball-pen and Langmuir probes measurements in COMPASS and found very good agreement between them. Revisiting the aforementioned JET CAV analysis indeed shows that the electron temperatures are much higher than previously reported, close to those predicted by the PIC simulation, and thus the ion energy seems to not significantly increase in the scrape-off layer., QC 20230425

Supported by the world magnetic fusion research community, a series of International Symposia on Liquid metals Applications for fusion (ISLA) have been held biannually since 2010. The 7th edition (ISLA-7) was held for the period from 12 December through 16 December 2022, at Chubu University located in Kasugai, Aichi, Japan. For the first time in the history of this series of symposia, ISLA-7 was held in a hybrid fashion, due to the COVID-19 situation. The total number of the participants was 60, 34 out of whom attended the symposium in person, and the rest participated online. As to the presentation statistics, 29 papers were presented in person, whereas 21 presentations were delivered online but real-time by the presenters in China, Spain, the UK, and the USA. Both of the presentations delivered in person and online were recorded, and the video has been shared by all participants. These participants represent 11 countries: China, Czech, Italy, Japan, Latvia, Netherlands, Russia, Thailand, the UK, and the USA. All these numbers are among the largest in this series of symposia. Covered by these presentations are; in session-2, program overviews and liquid metal research review; in session-3, liquid metal flows, and MHD issues; in session-4, liquid metal facilities; in sessions-5 and 6, liquid metal experiments and modeling; in session-7, divertor physics and heat flux mitigation; in session-8, plasma and liquid metals interactions; in session-9 liquid metal plasma-facing components, erosion, and wettability. In addition, there were an opening session whereby several opening addresses were delivered and also a closing session where all technical session summaries were presented by the respective session chairs.

One of the ways to extend the lifetime of the divertor for DEMO could be to replace the solid tungsten plasma-facing components with liquid tin (Sn) confined in a tungsten capillary porous structure (CPS). Testing a CPS in a divertor plasma configuration is crucial for the development of a liquid metal divertor (LMD) to understand how the main plasma is affected. Only a limited Sn concentration is allowed in the plasma core, due to the high radiative losses associated with the high atomic number of Sn (50). Therefore, it is necessary to test a small-scale LMD filled with Sn in a tokamak environment, which has not previously been done. In ASDEX Upgrade, a liquid tin module (LTM) has been exposed by means of the divertor manipulator. During plasma flat-top, the outer strike point (OSP) was placed onto the pre-heated LTM and held there for a time interval between 2 and 3.4s over multiple discharges. Photographs of the LTM taken after each discharge, revealed macroscopic Sn leakage onto the adjacent tile. Simulations with the HeatLMD code predicted an acceptable tin erosion near the LTM with thermal sputtering dominating over evaporation. However, spectroscopic measurements revealed an order of magnitude higher erosion. Since this remained constant when the OSP was held on the LTM so that the surface temperature increased, evaporation could be excluded as the main source of Sn erosion. Comparison between discharges with different durations of OSP location on the LTM revealed an increase in core radiation up to 1.5MW due to Sn. The 1.5D-impurity transport code STRAHL was used to interpret this increase in total plasma radiation and revealed a Sn concentration in the main plasma of up to 1.4×10-4. Given that the LTM only covered about 1/650 of the outer divertor circumference, extrapolating to a full toroidal divertor implies erosion is above acceptable limits. The unexpectedly high Sn fraction in the main plasma is attributed to the ejection of Sn droplets

Using liquid metals confined in capillary porous structures (CPSs) as a plasma-facing component (PFC) could prolong the lifetime of the divertor in the high heat flux area. However, the high atomic number of tin (Sn) limits its acceptable fraction in the main plasma. Therefore, a crucial step in developing this concept is to test it in a tokamak environment, particularly in the diverted plasma region, e.g. ASDEX Upgrade (AUG). In this paper, the design of liquid tin module (LTM) is explained, and the testing in the high heat flux device GLADIS before its use in AUG is presented. The LTM was additively manufactured using selective laser melting, consisting of a 1.5mm porous layer tungsten (W) directly attached to a solid W bulk. The LTM has a plasma-facing area of 16×40mm2 and was filled with 1.54g of Sn. In GLADIS, the module was exposed to power loads between 2 and 8MWm−2 for 1 up to 10s, first unfilled and later filled with Sn. The surface temperature was monitored with infrared imaging and pyrometry. The thermal response was used to compare with simulations in Ansys Mechanical, enabling a determination of the module's effective thermal properties. Sn droplets could be observed on the infrared camera, until a surface temperature of about a 1000°C was reached. The enhanced wetting of tin on the plasma-facing surface, which was observed by a visible camera, suggests that there is a conditioning of the surface, possibly due to the removal of impurities and oxides. Subsequent examinations of the adjacent tile revealed minor Sn leakages emanating from the module's edge. Furthermore, the module showed no indication of mechanical failure. Therefore, these results indicated that the LTM qualifies for the heat fluxes expected in ASDEX Upgrade.

Liquid metal based divertors could be a more robust alternative to a solid tungsten design for DEMO. The liquid is confined in a sponge-like tungsten layer, called a capillary porous structure (CPS). It has been found previously that under certain conditions, many tin droplets eject from a CPS when it is brought into contact with a hydrogen plasma. These would present a contamination issue for the plasma core. Stability analysis suggests that droplet ejection can be suppressed by reduction of the pore size. To test this, stainless-steel CPS targets with pore size ranging from 0.5–100µm filled with tin were exposed to identical loading conditions. This was done in the linear plasma device Magnum-PSI, capable of reaching divertor relevant plasma conditions. Furthermore, the influence of the CPS manufacturing techniques is considered by comparing the performance of a 3D printed, a mesh felts and a sintered CPS, all made from tungsten. Each target was surrounded by four witness plates, which were analysed post-mortem for Sn content by Rutherford backscattering. During plasma exposure, tin droplets were observed using a fast visible camera and plasma light emission via survey optical emission spectroscopy. The results imply that Sn erosion can be reduced by a factor of 50 when reducing the pore size. Moreover, it highlights the importance of avoiding overfilling of CPS targets with Sn.

In this work, we provide the first 2D spatially resolved description of radiative detachment in MAST-U Super-X L-mode divertor plasmas. The Super-X magnetic configuration was designed to achieve reduced heat- and particle loads at the divertor target compared to conventional exhaust solutions. We use filtered camera imaging to reconstruct 2D emissivity profiles in the poloidal plane for multiple atomic and molecular emission lines and bands. A set of deuterium fuelling scans is discussed that, together, span attached to deeply detached divertor states observed in MAST-U. Emissivity profiles facilitate separate analysis of locked-mode induced split branches of the scrape-off layer. Molecular deuterium Fulcher band emission front tracking reveals that the deuterium electron-impact ionisation front, for which it serves a proxy, detaches at different upstream electron densities in the split branches. Upon detachment of this ionisation front, Balmer emission attributed to molecular activated recombination appears near-target. We report a simultaneous radial broadening of the emission leg, consistent with previous SOLPS-ITER modelling. With increased fuelling this emission region detaches, implying electron temperatures below ∼ 1 eV. In this phase, 2D Balmer line ratio reconstruction indicates an onset of volumetric direct electron-ion recombination near-target. At the highest fuelling rates this emission region moves off-target, suggesting a drop in near-wall electron density accompanying the low temperatures.

In some conditions, I-mode plasmas can feature pedestal relaxation events (PREs) that transiently enhance the energy reaching the divertor target plates. To shed light on their appearance, characteristics and energy reaching the divertor targets, a comparative study between two tokamaks — Alcator C-Mod and ASDEX Upgrade — is carried out. It is found that PREs appear only in a subset of I-mode discharges, mainly when the plasma is close to the H-mode transition. Also, a growing oscillating precursor before the PRE onset is observed in the region close to the separatrix in both devices, and a discussion on a possible triggering mechanism is outlined. The PRE relative energy loss from the confined region is found to increase with decreasing pedestal top collisionality ${\nu }_{\text{ped}}^{\ast }$. Similarly, also the relative electron temperature drop at the pedestal top, which is related to the conductive energy loss, rises with decreasing ${\nu }_{\text{ped}}^{\ast }$. Based on these relations, the PRE relative energy loss in future devices such as DEMO and ARC is estimated. Finally, the divertor peak energy fluence due to the PRE is measured on each device. Those values are then compared to the model introduced in Eich et al (2017 Nucl. Mater. Energy 12 84–90) for type-I edge localized modes. The model is shown to provide an upper boundary for PRE energy fluence data, while a lower boundary is found by dividing the model by three. These two boundaries are used to make projections of the PRE divertor energy fluence to DEMO and ARC.

We present recent highlights from the most recent operation phases of Wendelstein 7-X, the most advanced stellarator in the world. Stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, have been maintained for tens of seconds. Pellet fueling allows for plasma phases with reduced ion-temperature-gradient turbulence, and during such phases, the overall confinement is so good (energy confinement times often exceeding 200 ms) that the attained density and temperature profiles would not have been possible in less optimized devices, since they would have had neoclassical transport losses exceeding the heating applied in W7-X. This provides proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X plasmas generally show good impurity screening and high plasma purity, but there is evidence of longer impurity confinement times during turbulence-suppressed phases.

We present recent highlights from the most recent operation phases of Wendelstein 7-X, the most advanced stellarator in the world. Stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, have been maintained for tens of seconds. Pellet fueling allows for plasma phases with reduced ion-temperature-gradient turbulence, and during such phases, the overall confinement is so good (energy confinement times often exceeding 200 ms) that the attained density and temperature profiles would not have been possible in less optimized devices, since they would have had neoclassical transport losses exceeding the heating applied in W7-X. This provides proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X plasmas generally show good impurity screening and high plasma purity, but there is evidence of longer impurity confinement times during turbulence-suppressed phases.

Tritium (T) retention characteristics in dust collected from the divertor in JET with ITER-like wall (JET-ILW) after the third campaign in 2015-2016 (ILW-3) have been examined in individual dust particles by combining radiography (tritium imaging plate technique) and electron probe micro-analysis. The results are summarized and compared with the data obtained after the first campaign in 2011-2012 (ILW-1). The dominant component in ILW-1 dust was carbon, QC 20220308

In the FFHR power reactor equipped with a supercritical CO2 gas turbine power generation system, an divertor cooling system is connected to this power generation system [S. Ishiyama et al., Prog. Nucl. Energy 50, No.12-6, 325 (2008) [1]]. In this paper, for the purpose of developing a diverter by supercritical CO2 gas cooling that can cope with a neutron heavy irradiation environment with a heat load of 15 MW/m2 or more, CFD heat transfer flow analysis was carried out for performance evaluation and its design optimization by a structural analysis models of a supercritical CO2 gas cooled divertors. As a result, in the supercritical CO2 gas cooled tungsten mono-block divertors (50 × 50 mm × 5 channel × 5,000 mL) with a flow path length of 5 m or less, the engineering designable range of these advanced diverters having the same cooling performance as the water cooling divertor was clarified, and its practicality is extremely high from the feature that the structural model has an extremely low risk during operation as compared with the water cooled divertor., source:https://doi.org/10.1585/pfr.17.1405103

An approach to describe the plasma parameter variation with the effective minor radius across the edge region in the heliotron Large Helical Device (LHD) in configurations without and with a resonant magnetic perturbation (RMP) is elaborated, by averaging fluid equations for transport of particles, momentum and heat over the flux surfaces. Numerical solutions of one-dimensional, time-dependent equations derived and analytical estimates performed allow us to interpret the LHD experiments on the density limit. Calculations reproduce qualitatively the principal difference between situations without and in the presence of RMP: in the former case a thermal collapse of the discharge occurs immediately after the plasma detachment from the divertor target plates at a radiation level of 0.4–0.5; in the latter the radiating layer is localized at the plasma edge even if the power radiated exceeds 90% of the input power. A prominent role for such a behavior of plasma particle flows along magnetic field lines perturbed by RMP, leading to a positive radial gradient of the plasma density inside the magnetic island, is demonstrated., source:Mikhail Tokar et al 2020 Plasma Phys. Control. Fusion 62 085011, source:https://doi.org/10.1088/1361-6587/ab9c49, identifier:0000-0002-9536-9841

National Institute for Fusion Science (NIFS) launched in 2014 a research program for developing Dispersion Strengthened (DS) Cu alloys for application to the heat sink materials of divertors of fusion reactors, using newly installed ball-milling, encapsulation, and Hot Isostatic Pressing (HIP) facilities. A unique feature of these facilities is that the entire process can be performed without exposing the materials to air, enabling precise impurity control. Cu-Al, Cu-Zr and Cu-Y alloys have been produced in this program. Various technological advancement has been made for the fabrication, such as suppression of powder adhesion to the wall of containers during the ball milling, and encapsulation technology including development of small volume tubular capsules., source:http://dx.doi.org/10.4028/www.scientific.net/MSF.941.778

The required pumping power for the liquid metal ergodic limiter/divertor REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor), which is considered as an optional divertor design for the LHD-type helical fusion reactor FFHR-d1, has been estimated with a consideration of the change in the strength of transverse magnetic field along the flow channel. It is found that the pressure drop and the required pumping power can be suppressed to an acceptable level by using insulated ducts., source:https://doi.org/10.1585/pfr.12.1405016, identifier:0000-0001-6028-8980

In this study, we newly propose two types of water-cooled divertor devices called EVAPORON-3 using high thermal conductivity porous media that could be applicable especially to an enlarged heat transfer surface such as the divertor plate. Each device has a liquid-vapor separating plate (LVS plate) on the top surface of the porous medium. The LSV plate of the Type-1 device has 9 inlet holes for liquid supply and 12 outlet holes for vapor discharge against a heat transfer surface with the diameter of 30 mm. On the other hand, the LSV plate of the Type-2 device has just one liquid inlet hole at the center of the plate and 12 outlet holes for the vapor discharge. The diameter of each hole is 2.6 mm. The introduced porous medium is a particles-packed bed with the particle diameter of 1.0 mm. In order to reduce pressure loss of the liquid and vapor flows in the porous medium, two layers of the particle are set between the heat transfer surface and the LSV plate. The thickness of the porous bed is 1.82 mm. The result showed that each device enabled a high heat removal of over 5 MW/m2 and, in particular, the Type-2 device succeeded in the heat flux removal of approximately 11 MW/m2 at the water flow rate of 2 L/min, although the pressure loss increased by decreasing the number of the inlet holes., source:https://doi.org/10.1585/pfr.12.1405015

Recently, a new limiter concept with liquid metal for a helical fusion reactor was proposed. This work present a modeling study to understand the influence of the limiter to the plasma and to find critical parameters for the liquid limiter concept. EMC3-EIRENE code is applied to a limiter configuration of a vertical column formed with plane surfaces at the inboard side of the Large Helical Device (LHD). The calculation results indicate that heat load on the helical divertor plates is significantly reduced when the limiter is inserted into the ergodic layer even without impurities. However, the influence of impurity radiation is too large in some cases. In order to achieve a large reduction of divertor flux and to avoid a large reduction of core electron temperature, parameter scans with limiter position and sputtering yield were performed. The results suggest that the limiter position is a critical parameter for acceptable degradation of the core and sufficient reduction of divertor flux., source:https://doi.org/10.1585/pfr.13.3403034, identifier:0000-0002-9117-5898

The stabilization effect of the Internal Flow Resistances (IFR) inserted into water jets has been investigated in order to apply the liquid metal divertor to the helical fusion reactor, FFHR. A high heat load exceeding 20 MW/m2 is one of the important issues for the divertor in the nuclear fusion reactor. A new concept of ergodic limiter/divertor called the REVOLVER-D has been proposed for FFHR. This divertor is expected to tolerate a high heat load larger than a few tens of MW/m2. Jets narrow after acceleration by gravity and transform to droplets due to the surface tension instability. Stabilization of jets is an important issue for realizing the REVOLVER-D. In this paper, stabilization of jet by various IFR has been examined. Here, the “sheath jet” is defined as a jet with an IFR inside. Experiments on sheath jet have been carried out and the stabilization effect of IFR has been confirmed. Numerical simulations on sheath jet using ANSYS have been carried out. It is shown that the velocity of the sheath jet reaches the terminal velocity, which is proportional to 0.36 ± 0.01 power of the flow rate and −0.26 ± 0.01 power of the outer circumference of the IFR., source:https://doi.org/10.1585/pfr.13.1405003, identifier:0000-0001-5079-6391

A fusible metal pebble divertor system named the REVOLVER-D2 is newly proposed here, based on the molten tin jet divertor REVOLVER-D and the pebble divertor concepts. In the REVOLVER-D2, solid tin pebble flows are used as the divertor target. Molten tin pools receive the falling tin pebbles to mitigate the impact of the heavy metal shower. All pebbles are melted in the molten tin pool and recycled into the tin pebbles again. The solid tin pebbles are formed inside the vacuum vessel by the shot tower method using silicone oil pools. This concept enables simultaneous achievement of a high heat load tolerance and a high evacuation efficiency, while mitigating the MHD effects. A high heat load can be removed by the pebble flow if the pebble size is large enough and the velocity is fast enough as with the molten tin jet. Although the charge up of pebbles affects the pebble motion due to the Lorentz force and the Coulomb force, the Lorentz force is negligible if the injection velocity is fast enough. For example, center-to-center pebble distance due to the Coulomb force will be ∼ 4 mm in the case of the injection velocity of 10 m/s if the molten tin pool is placed 2 m under the point where pebbles enter the ergodic layer., source:https://doi.org/10.1585/pfr.14.3405050, identifier:0000-0001-5079-6391

The realization of advanced fusion reactors rests upon improvements of plasma-facing materials for divertors. The toughness of tungsten is a critical issue for these improvements. As a solution, we developed dispersion-strengthened (DS) W-Ti alloys using a new process which is based on a combination of mechanical alloying (MA) and hot isostatic pressing. In this paper, the microstructure of new DS-W-Ti alloys and the formation mechanism of nano-strengthening particles are investigated. The produced DS-W-Ti alloys are dispersed with TiO2. The chemical analysis shows that new precipitation processes occurred during the MA based on redox reactions between carbon and oxygen., source:https://doi.org/10.1585/pfr.15.1205021, identifier:0000-0001-9016-2318

Liquid Sn is one of the promising coolants for liquid surface divertor concept of fusion reactors. However, the compatibility between liquid Sn and structural materials is an important issue that has to be addressed, because liquid Sn is extremely corrosive to steels at high temperatures. The corrosion may be mitigated when a protective Al2O3 layer is formed on the surface of alumina forming steels. However, the effect of neutron irradiation on the integrity of protective layer is not made clear so far. Japan and US joint research project “FRONTIER” started in 2019 to investigate the material compatibility under neutron irradiation. The purpose of the present study is to develop the conceptual design of the irradiation test capsule which enables material compatibility tests for the alumina forming steels - liquid metal systems under neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory, TN, USA. The three dimensional drawing of capsule structure was then developed. The validity of the material selections for the capsule design was investigated by means of corrosion tests of SiC, Si3N4, Ti, and Mo in liquid Sn at 773 K for 262 hr., source:https://doi.org/10.1585/pfr.16.2405040

A new divertor concept REVOLVER-D2 that uses metal pebbles as a divertor target has been proposed for the LHD-type helical fusion reactor FFHR-c1. Demonstration of sufficient plasma shielding rate is one of the important issues to show the feasibility of the REVOLVER-D2. For this purpose, an experiment on the plasma irradiation to the metal pebble flow has been conducted in the sheet plasma device TPDsheet-U in Tokai University. Reduction of the ion flux and electron density in the metal pebble flow was observed and it was confirmed that the metal pebble flow can play the role of the divertor target. The measured shielding rate is consistent with the prediction by the model based on the Lambert-Beer-Bouger’s law. Achievement of the shielding rate of > 72 % is expected in the FFHR-c1 using the nozzle with the size of > 310 mm × 310 mm and the pebble flow with the diameter of 5 mm., source:Takeru Ohgo, Takuya Goto, Toshikio Takimoto, Akira Tonegawa, Junichi Miyazawa, Plasma irradiation experiment on the metal pebble flow in the TPDsheet-U, Fusion Engineering and Design, Volume 165, 2021, 112236, ISSN 0920-3796, https://doi.org/10.1016/j.fusengdes.2021.112236., source:https://doi.org/10.1016/j.fusengdes.2021.112236

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 13C-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., source:J.P. Coad, M. Rubel, J. Likonen, N. Bekris, S. Brezinsek, G.F. Matthews, M. Mayer, A.M. Widdowson, Material migration and fuel retention studies during the JET carbon divertor campaigns, Fusion Engineering and Design, Volume 138, 2019, Pages 78-108, ISSN 0920-3796, https://doi.org/10.1016/j.fusengdes.2018.10.002., source:https://doi.org/10.1016/j.fusengdes.2018.10.002

Long pulse discharges in the Large Helical Device have often been interrupted by large amounts of dust particle emission from the divertor region caused by the exfoliation of carbon-rich mixed material deposition layers. The plasma wall interaction code ERO2.0 has provided the simulation results of the three-dimensional distribution of the carbon flux density in the divertor region which is quite reasonable with the observed distribution of the carbon-rich deposition layers. The code has also succeeded in reproducing the reduction of the carbon deposition layers on dome plates by changing the target plate configuration in the divertor region. The ERO2.0 simulations have also successfully explained dust particle emission from the inboard side near the equatorial plane for the new target plate configuration at the termination of a long pulse discharge. These simulation results prove that the ERO2.0 code is applicable to predicting the possible position from where the dust particles are released, and to designing an optimized divertor configuration for performing stable long pulse discharges with controlled dust particle emission., source:https://doi.org/10.1585/pfr.16.2403004, identifier:0000-0003-0655-7347

When an ion or atom is injected into a target material, the incident energy increases the temperature around the injection point. Owing to the energy transfer to the surrounding atoms, the temperature of the material recovers to the initial temperature in the real system because heat is transferred to the bulk of material which consists of an almost infinite number of atoms. However, if we simulate the system by the molecular dynamics (MD) method, it is difficult to prepare a sufficiently large system to calculate the process of heat transfer to the bulk. Therefore, in this study, we develop an MD simulation model which includes the effect of heat transfer to the bulk by solving heat conduction equation. The simulation result shows that the process of heat removal proceeds on a time scale of approximately 1000 times slower than the time scale of the incident energy transfer from the incident atom to the target material., source:https://doi.org/10.1585/pfr.15.2403073

The 3D peripheral plasma and neutral transport code, EMC3-EIRENE was applied to the Heliotron J with a wide and flexible controllability of magnetic configuration. This code requires a three-dimensional (3D) grid with high resolution in the peripheral plasma region to reproduce the fine plasma structure. The grid generation tool, FLARE, was utilized to create the grid in conjunction with the code developed to arbitrarily set the outer boundary of the peripheral grid. After setting up the 3D grid, we carried out the EMC3-EIRENE calculation successfully for the first time in the Heliotron J's standard configuration. In addition, the convergence of the iterative calculation and the effects of different grid resolutions upon the calculation were investigated. A good numerical convergence was obtained, and the influence of resolution was observed in the electron density in the divertor region., source:https://doi.org/10.1585/pfr.14.3403127, identifier:0000-0002-5730-4158

Impurity transport in the peripheral plasma in the Large Helical Device for carbon and tungsten divertor configurations is analyzed using a three-dimensional edge plasma simulation code (EMC3-EIRENE). Long pulse plasma discharges in LHD have often been interrupted by the emission of iron dust from the surface of the vacuum vessel. The iron ions in the peripheral plasma originating from the iron dust can enhance the sputtering of impurities on the divertor plates. The influence on the peripheral plasma by the sputtered impurities is investigated in the iron dust emission case for both the carbon and the tungsten divertor configurations. The simulation reveals that the dependence of the radiation power by the sputtered impurities on the plasma density and the iron emission rate is quite different between the two divertor configurations. It is found that, for the tungsten divertor, a high plasma density operation is advantageous for controlling the radiation power and the impurity ion content. In the high plasma density operation, the radiation power by tungsten ions is suppressed to less than that by carbon ions for the carbon divertor even in the case of high iron dust emission rates due to the reduced sputtering rate by lowered divertor electron temperatures., source:https://doi.org/10.1585/pfr.14.3403057, identifier:0000-0003-0655-7347

In order to understand plasma transport in the edge stochastic layers, a visible spectrometer has been developed. Two-dimensional impurity emission intensity distributions in the edge stochastic layer of LHD, including divertor plate, divertor legs, X-point, and the last closed flux surface, have been measured. Carbon emission intensities and their distributions are found to be clearly different depending on the different magnetic field configuration and on the charge state. Effects of the edge magnetic field structure on these results are discussed., source:https://doi.org/10.1585/pfr.13.3402030, identifier:0000-0002-7368-8614

In order to study migration of sputtered materials and profiles of their re-deposition in the Large Helical Device (LHD), material probes fixed with Si specimens were installed on the private region of the torus outer first-wall in all toroidal sections (10 sections) in the experimental campaign in 2011. The material probes fixed with the Si specimens were exposed to various wall conditionings such as glow discharges cleanings (GDCs), boronization, Ti sublimation and main plasma discharges. After the campaign, the specimens surfaces were investigated with visual observation, Scanning Electron Microscope (SEM) and Energy Dispersive X-ray Spectroscopy (EDS). A type of visual deposition (band) pattern has been observed on the probes at almost all toroidal sections of LHD. The analysis of these patterns on the material probes informs the source and direction of material migration inside the LHD vacuum vessel. The observation and result of the experiment have been presented in this paper., source:https://doi.org/10.1585/pfr.8.2402074

A joint of oxide dispersion strengthened copper alloy (ODS-Cu; GlidCop®), i.e., GlidCop®/GlidCop® joint and a joint of stainless steel (SUS) and GlidCop®, i.e., SUS/GlidCop® joint were prepared by using the advanced brazing technique (ABT). Deformation and fracture behavior, and thermal stability of these joints were investigated by means of three-point bending test, scanning electron microscope (SEM) and energy dispersive X-ray spectroscopy (EDS). Firstly, in the case of the GlidCop®/GlidCop® joint, plastic deformation occurred predominantly in the diffusion layer with lower hardness. The fracture proceeded approximately at the center of the diffusion layer. In the case of the SUS/GlidCop® joint, on the other hand, plastic deformation occurred widely in the bulks, GlidCop® and SUS near the interfaces between the bulk and the bonding layer. The fracture proceeded mainly along the interface between the bonding layer and the bulk GlidCop®, and substantially along the interface between the bonding layer and the bulk SUS. Secondly, influences of 2nd-step heat on the GlidCop®/GlidCop® and SUS/GlidCop® joints were examined from viewpoints of microstructure and mechanical strength. Almost no obvious changes in the width of the diffusion layer (GlidCop®/GlidCop® joint) or the bonding layer (SUS/GlidCop® joint), and in the element distribution in these layers after the 2nd-step heat were recognized. In addition, any obvious changes in the shape of stress-strain curve and yield strength were not recognized for both joints., source:https://doi.org/10.1016/j.fusengdes.2022.113312, identifier:0000-0002-3744-2481

We present recent highlights from the most recent operation phases of Wendelstein 7-X, the most advanced stellarator in the world. Stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, have been maintained for tens of seconds. Pellet fueling allows for plasma phases with reduced ion-temperature-gradient turbulence, and during such phases, the overall confinement is so good (energy confinement times often exceeding 200 ms) that the attained density and temperature profiles would not have been possible in less optimized devices, since they would have had neoclassical transport losses exceeding the heating applied in W7-X. This provides proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X plasmas generally show good impurity screening and high plasma purity, but there is evidence of longer impurity confinement times during turbulence-suppressed phases., This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 and 2019-2020 under Grant Agreement No. 633053., Peer Reviewed, Article signat per 497 autors/es: Thomas Sunn Pedersen1,2,∗ , I. Abramovic3, P. Agostinetti4, M. Agredano Torres1, S. Äkäslompolo1, J. Alcuson Belloso1, P. Aleynikov1, K. Aleynikova1, M. Alhashimi1, A. Ali1, N. Allen5, A. Alonso6, G. Anda7, T. Andreeva1, C. Angioni8, A. Arkhipov8, A. Arnold1, W. Asad8, E. Ascasibar6, M.-H. Aumeunier9, K. Avramidis10, E. Aymerich11, S.-G. Baek3, J. Bähner1, A. Baillod12, M. Balden1, M. Balden8, J. Baldzuhn1, S. Ballinger3, M. Banduch1, S. Bannmann1, A. Banon Navarro8, A. Bañon Navarro ´ 1, T. Barbui13, C. Beidler1, C. Belafdil9, A. Bencze7, A. Benndorf1, M. Beurskens1, C. Biedermann1, O. Biletskyi14, B. Blackwell15, M. Blatzheim1, T. Bluhm1, D. Böckenhoff1, G. Bongiovi16, M. Borchardt1, D. Borodin17, J. Boscary8, H. Bosch1,18, T. Bosmann19, B. Böswirth8, L. Böttger1, A. Bottino8, S. Bozhenkov1, R. Brakel1, C. Brandt1, T. Bräuer1, H. Braune1, S. Brezinsek17, K. Brunner1, S. Buller1, R. Burhenn1, R. Bussiahn1, B. Buttenschön1, A. Buzás7, V. Bykov1, I. Calvo6, K. Camacho Mata1, I. Caminal20, B. Cannas11, A. Cappa6, A. Carls1, F. Carovani1, M. Carr21, D. Carralero6, B. Carvalho22, J. Casas20, D. Castano-Bardawil17, F. Castejon6, N. Chaudhary1, I. Chelis23, A. Chomiczewska24, J.W. Coenen13,17, M. Cole1, F. Cordella25, Y. Corre9, K. Crombe26, G. Cseh7, B. Csillag7, H. Damm1, C. Day10, M. de Baar27, E. De la Cal6, S. Degenkolbe1, A. Demby13, S. Denk3, C. Dhard1, A. Di Siena8,28, A. Dinklage12, T. Dittmar17, M. Dreval14, M. Drevlak1, P. Drewelow1, P. Drews17, D. Dunai7, E. Edlund3, F. Effenberg29, G. Ehrke1, M. Endler1, D.A. Ennis5, F.J. Escoto6, T. Estrada6, E. Fable8, N. Fahrenkamp1, A. Fanni11, J. Faustin1, J. Fellinger1, Y. Feng1, W. Figacz4, E. Flom13, O. Ford1, T. Fornal24, H. Frerichs13, S. Freundt1, G. Fuchert1, M. Fukuyama30, F. Füllenbach1, G. Gantenbein10, Y. Gao1, K. Garcia13, J.M. García Regaña6, I. García-Cortés6, J. Gaspar31, D.A. Gates29, J. Geiger1, B. Geiger13, L. Giudicotti32, A. González6, A. Goriaev26,33, D. Gradic1, M., Objectius de Desenvolupament Sostenible::7 - Energia Assequible i No Contaminant, Objectius de Desenvolupament Sostenible::7 - Energia Assequible i No Contaminant::7.a - Per a 2030, augmentar la cooperació internacional per tal de facilitar l’accés a la investigació i a les tecnolo­gies energètiques no contaminants, incloses les fonts d’energia renovables, l’eficiència energètica i les tecnologies de combustibles fòssils avançades i menys contaminants, i promoure la inversió en infraestructures energètiques i tecnologies d’energia no contaminant, Postprint (published version)

We present recent highlights from the most recent operation phases of Wendelstein 7-X, the most advanced stellarator in the world. Stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, have been maintained for tens of seconds. Pellet fueling allows for plasma phases with reduced ion-temperature-gradient turbulence, and during such phases, the overall confinement is so good (energy confinement times often exceeding 200 ms) that the attained density and temperature profiles would not have been possible in less optimized devices, since they would have had neoclassical transport losses exceeding the heating applied in W7-X. This provides proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X plasmas generally show good impurity screening and high plasma purity, but there is evidence of longer impurity confinement times during turbulence-suppressed phases.

Self-consistent full-size turbulent-transport simulations of the divertor and scrape-off-layer (SOL) of existing tokamaks have recently become feasible. This enables the direct comparison of turbulence simulations against experimental measurements. In this work, we perform a series of diverted ohmic L-mode discharges on the tokamak à configuration variable (TCV) tokamak, building a first-of-a-kind dataset for the validation of edge turbulence models. This dataset, referred to as TCV-X21, contains measurements from five diagnostic systems from the outboard midplane (OMP) to the divertor targets - giving a total of 45 one- and two-dimensional comparison observables in two toroidal magnetic field directions. The experimental dataset is used to validate three flux-driven 3D fluid-turbulence models - GBS, GRILLIX and TOKAM3X. With each model, we perform simulations of the TCV-X21 scenario, individually tuning the particle and power source rates to achieve a reasonable match of the upstream separatrix value of density and electron temperature. We find that the simulations match the experimental profiles for most observables at the OMP - both in terms of profile shape and absolute magnitude - while a comparatively poorer agreement is found towards the divertor targets. The match between simulation and experiment is seen to be sensitive to the value of the resistivity, the heat conductivities, the power injection rate and the choice of sheath boundary conditions. Additionally, despite targeting a sheath-limited regime, the discrepancy between simulations and experiment also suggests that the neutral dynamics should be included. The results of this validation show that turbulence models are able to perform simulations of existing devices and achieve reasonable agreement with experimental measurements. Where disagreement is found, the validation helps to identify how the models can be improved. By publicly releasing the experimental dataset and validation analysis, this wo

The capabilities of the newly installed divertor Thomson Scattering (DTS) diagnostic (Kurzan et al 2021 J. Inst. 16 C09012) at ASDEX Upgrade (AUG) have been demonstrated by measuring 2D electron densities n e and temperatures T e from attached to fully detached divertor conditions in L- and H-mode. The collected dataset is a breakthrough for divertor studies at AUG in which such measurements have been so far missing. Besides highlighting the strengths and limits of the DTS system, this work provides confirmations and new insights into detachment physics. The transition between partial and pronounced detachment correlates with a 2D redistribution of the electron density from the inner to the outer divertor in both L- and H-mode. In pronounced detachment, a strong parallel pressure gradient could be confirmed throughout the complete SOL. Finally, measurements of n e and T e of a stable X-point radiator revealed local temperatures as low as 1 eV within the confined plasma, whereas the pressure is shown to be constant on closed field lines.

©2021 Elsevier The present work has been supported by the Agencia Estatal de Investigaci ' on (PID2019-105325RB-C33/AEI/10.13039/501100011033) and by the Regional Government of Madrid through the program TECHNOFUSION(III)CM (S2018/EMT-4437). The support of the Regional Government of Madrid through the multi-annual agreement with UC3M ("Excelencia para el Profesorado Universitario" EPUC3M14) -Fifth regional research plan 2016-2020 is acknowledge., Dispersion strengthened copper alloys have been produced following an innovative powder metallurgy route. Copper and yttrium acetate powders have been mechanically alloyed and posteriorly thermal treated at 923 K for 3 h and 15 h under a hydrogen atmosphere in order to transform the yttrium acetate into Y_2O_3. Subsequently, the powders were consolidated by hot isostatic pressing. It has been concluded that the duration of the thermal treatment of the powder is a determining factor in the degree of densification of the alloy. The study of the microstructure by Scanning Electron Microscopy and Electron Backscatter Diffraction has revealed the presence of micrometer and submicrometer grains and nanometric Y-O enriched Cu particles embedded in the copper matrix, the mean grain size being smaller for the sample produced from the powder thermal treated for 15 h. Transmission Electron Microscopy investigations concluded that the nanoparticles exhibit a spherical shape with a size up to 25 nm and correspond to monoclinic Y_2O_3. Annealing twins have been also observed, especially in the material produced from thermal treated powder for longer. The mechanical properties have been inferred from Vickers microhardness measurements and compression tests. Below 473 K the yield strengths of the produced materials are greater than that of pure copper and above 473 K are close to them. From the study of the thermal properties of the densest material it has been found that its thermal conductivity remains nearly constant in the temperature range 300-773 K, and its value is around 85% the thermal conductivity of CuCrZr, the reference material for ITER., Ministerio de Ciencia e Innovación (MICINN), Comunidad de Madrid, Depto. de Física de Materiales, Fac. de Ciencias Físicas, TRUE, pub

[ES] ASDEX Upgrade (AUG) es un Tokamak experimental (www.aug.ipp.mpg.de) en funcionamiento desde 1991. Se trata de un Tokamak de tamaño medio diseñado para investigación de la energía de escape con una potencia de calentamiento normalizada relevante para el ITER P/R de 18 MW/m. A lo largo de los años, se han realizado muchas actualizaciones y en los próximos años se instalará un nuevo divertor superior con bobinas dentro de la vasija. El objetivo del proyecto es demostrar que configuraciones alternativas (¿X-divertor¿ o ¿Snowflake divertor¿) pueden mitigar el problema del a energía de escape. Para obtener las configuraciones magnéticas requeridas, se deben integrar en el diseño del divertor dos bobinas de campo poloidal cerca de la línea de ataque. Esta mejora se llevará a cabo en el divertor superior del AUG que, desde su construcción, nunca ha sido modificado. Por lo tanto, en 2022 se instalarán nuevos componentes: la bomba criogénica, los divertors internos y externos y las bobinas internas. El diseño del divertor exterior está fuertemente impulsado por muchos requisitos impuestos por la física: el más severo es la integración de las bobinas en los módulos del divertor; otro es la necesidad de helicidad en las configuraciones del campo magnético. Para combinar ambas restricciones, el divertor exterior está diseñado como un anillo rígido formado por módulos bridados entre sí. De esta manera, no es necesario inclinar los módulos ni los elementos objetivo se fijados a ellos. Las placas objetivo están hechas de grafito revestido de tungsteno y están montadas en la estructura del anillo toroidal de soporte situado en el interior de la vasija de vacío. Una vez que los objetivos estén instalados, tendrán un espacio toroidal de 1 mm y un paso de ±0,1 mm entre los objetivos adyacentes. La forma de las placas está optimizada para reducir el efecto de los bordes de ataque que podrían terminar en un factor de amplificación del flujo de calor. A lo largo del perfil poloidal d, [EN] ASDEX Upgrade (AUG) is an experimental Tokamak (www.aug.ipp.mpg.de) in operation since 1991. It is a midsize Tokamak designed for power exhaust research with an ITERrelevant normalized heating power P/R of 15 MW/m. Over the years, many upgrades have been realized and in the next years a new upper divertor with in-vessel coils will be installed. The aim of the project is to prove that alternative divertor configurations (X-divertor or Snowflake divertor) can mitigate the exhaust power problem. To realize the required magnetic configurations, two poloidal field coils close to the strike line must be integrated in the design of the divertor. This upgrade will be carried out in the upper divertor of AUG that, since its building, has never been modified. Therefore, new components are going to be installed in 2022: cryopump, inner and outer divertor targets, and the in-vessel coils. The design of the outer divertor is strongly driven by the integration of the coils in the divertor modules and necessity of helicity in the magnetic field configurations. To combine both, the outer divertor is designed as a stiff ring consisting of modules flanged together that do not need to be tilted. The target elements are made of graphite coated with tungsten and they are mounted inside the vacuum vessel in the already existing toroidal ring structure. The shape of the targets is optimized to reduce the effect of leading edges that cause an increment of the heat flux due to the orientation of the smashing particles The outer upper divertor has three types of target design. The one expecting the higher heat flux is designed without front mounting holes, to avoid any edge effect. The maximum expected heat flux power is in the order of 10 MW/m2 and the targets are weakly cooled. They receive the energy during the 10 s long pulse and are cooled down via the clamping to the water-cooled support structure in between discharges. This master thesis is focused on the evaluation of the therma

Detachment experiments have been carried out in the linear plasma device Magnum-PSI by increasing the gas pressure near the target. In order to have a proper detailed analysis of the mechanism behind momentum and power loss in detachment, a quantitative match is pursued between B2.5-Eunomia solutions and experimental data. B2.5 is a multi fluid plasma code and Eunomia is a Monte Carlo solver for neutral particles, and they are coupled together to provide steady-state solution of the plasma and neutral distribution in space. B2.5-Eunomia input parameters are adjusted to produce a close replication of the plasma beam measured in the experiments without any gas puffing in the target chamber. Using this replication as an initial condition, the neutral pressure near the plasma beam target is exclusively increased during simulation, matching the pressures measured in the experiments. Reasonable agreement is found between the electron temperature of the simulation results with experimental measurements using laser Thomson scattering near the target. The simulations also reveal the effect of increased gas pressure on the plasma current, effectively reducing the current penetration from the plasma source. B2.5-Eunomia is capable of reproducing detachment characteristics, namely the loss of plasma pressure along the magnetic field and the reduction of particle and heat flux to the target. The simulation results for plasma and neutrals will allow future studies of the exact contribution of individual plasma-neutral collisions to momentum and energy loss in detachment in Magnum-PSI.

Liquid metal (LM) divertors are considered for the European DEMO reactor, because they may offer improved performance compared to the tungsten monoblock concept. The goal of this work is to provide a concept design, and explore the limitations of liquid metal divertors. To this end, a set of design requirements was formulated in close collaboration with the EUROfusion Power Plant Physics and Technology team (responsible for the design of the EU-DEMO). Tin was chosen as the preferred liquid metal, because unacceptable Tritium retention issues arise when lithium is used in DEMO. A concept design was then chosen that consists of water cooled pipes that are square on the outside and round on the inside, a corrosion barrier, and a 3D-printed porous tungsten armor layer filled with liquid tin. The porous armor layer acts as a Capillary Porous System (CPS). The design was analyzed using thermo-mechanical FEM simulations for various armor thicknesses and heat sink materials: Densimet, W/Cu composites, and CuCrZr. The highest heat loading capability achieved is 26.5 MW/m2 in steady state (18.9 MW/m2 when taking into account a safety margin of 1.4). This is achieved using a CuCrZr pipe, with a 1.9 mm thick armor. When increasing the armor layer to 3 mm thick, more than 80 MW/m2 can be withstood during slow transients thanks to vapor shielding, but at the same time the steady-state capability is reduced to 18 MW/m2. Resilience against disruptions cannot yet be proven, but is deemed within the realm of possibility based on estimates regarding the behavior of vapor shielding. This should be further investigated. Overall, the concept is considered a significant improvement compared to the original specifications (which are also the specifications to the tungsten monoblocks: 10 MW/m2 in steady state, and ∼20 MW/m2 during slow transients). Moreover, the possibility of withstanding disruptions is regarded as a pote

The Tokamak à Configuration Variable (TCV) has recently been equipped with gas baffles to increase its divertor closure for a broad range of divertor magnetic geometries. First experimental results reported in Reimerdes et al. (2021) demonstrated compatibility with a broad range of divertor magnetic geometries and confirmed the main design constraints of the baffles, in particular an increased divertor neutral pressure. The present article presents a more in-depths analysis and extended experiments of this first baffle assessment on the TCV boundary plasma. It is shown that the divertor neutral pressure increased following the installation of the baffles, as predicted by SOLPS-ITER and SolEdge2D-EIRENE simulations. Varying the divertor closure by changing the position of the plasma showed that the plasma equilibrium designed to assess the baffle effect was not far from the optimal trade-off between plasma plugging and recycling on the baffles. The baffles facilitate access to a partially detached regime in both L- and H-modes. In L-Mode, with the ion ∇B-drift directed from the X-Point to the plasma core, a reduction of the line-averaged density detachment threshold by approximately 20% is observed at the outer target, while inner strike point detachment is only achieved in the presence of baffles. Multispectral imaging shows that the CIII front moves from the outer target towards the X-Point at a lower line-averaged plasma density, indicating a colder outer leg. In H-mode, the CIII front is generally located near the X-Point between the ELMs, while without baffles, N2-seeding is required to move the front up to that location.

Balancing exploration and anti-predation are fundamental to the fitness and survival of all animal species from early life stages. How these basic survival instincts drive learning remains poorly understood. Here, using a light/dark preference paradigm with well-controlled luminance history and constant visual surrounding in larval zebrafish, we analyzed intra- and intertrial dynamics for two behavioral components, dark avoidance and center avoidance. We uncover that larval zebrafish display short-term learning of dark avoidance with initial sensitization followed by habituation; they also exhibit long-term learning that is sensitive to trial interval length. We further show that such stereotyped learning patterns is stimulus-specific, as they are not observed for center avoidance. Finally, we demonstrate at individual levels that long-term learning is under homeostatic control. Together, our work has established a novel paradigm to understand learning, uncovered sequential sensitization and habituation, and demonstrated stimulus specificity, individuality, as well as dynamicity in learning.