Metal Divertor Research Articles

Scaling of HeatLMD-simulated impurity outflux from COMPASS-U liquid metal divertor

Abstract The liquid metal divertor (LMD) concept offers a promising solution to manage extreme heat loads in plasma devices. This study presents predictive simulations using the HeatLMD model for the COMPASS-U tokamak with a full toroidal liquid metal divertor, expected to achieve reactor-relevant divertor heat flux densities. We derive the scaling of the Li|Sn outflux over 7 assumed independent parameters, transferable to other tokamaks. Its transport to LCFS (via ERO2.0) and its radiation (via Aurora and FACIT) predicts acceptably low lithium concentration and negligible plasma cooling. However, for tin, the medium power scenario requires backside cooling beyond the capability of the ITER-like water-cooled divertor, though a temporary heat absorber can approximate this for a 1 s plasma pulse. For incident divertor power exceeding 2 MW and strike point T e < 10 eV, HeatLMD predicts significant tin plasma radiative disruption.

Stabilization of liquid tin surfaces under D plasma irradiation by a capillary porous system

Liquid metal divertors are discussed as a potential alternative to divertor components made entirely of solid materials such as tungsten. However, liquid metal interaction with H isotope plasmas poses entirely new challenges, including the stability of the liquid surface. In this article, the interaction of low-temperature D plasmas with both free liquid tin surfaces, and the liquid metal confined in a capillary porous system (CPS) is presented. Free surfaces showed intolerable instabilities due to formation of gas bubbles in the melt, as opposed to a strong stabilizing effect of the CPS. A tendency of liquid tin to well up and entirely cover the surface of the CPS may be beneficial in terms of quickly re-establishing a tin surface after depletion by transient overloading. On the other hand, this appears to counteract efforts to fully suppress superficial bubble formation and bursting, and the resulting ejection of tin droplets by designing a CPS with very small pores.

Investigation of the spreading of a liquid metal droplet under a vertical magnetic field

In the liquid metal divertor of a magnetic confinement fusion device, the spreading characteristics of the liquid metal are crucial for ensuring the stable operation of the divertor. This study has experimentally investigated the spreading characteristics of a GaInSn alloy droplet on a solid substrate under a strong vertical magnetic field, with the magnetic field intensity ranging from 0 to 2.5 T. First, several parameters of the droplet, such as droplet shape, spreading factor, dynamic contact angle, spreading velocity, and rebound behavior after impacting, were studied without a magnetic field. The fitting relationship between maximum spreading factor βmax and Weber number We was obtained and has been compared with the scaling laws from the literature. Furthermore, the effect of the vertical magnetic field on those parameters has been investigated systematically. Quantitative results on βmax and the maximum spreading time tDmax, varied with the Hartmann number (Ha) and the We number, provide a comprehensive understanding of the spreading dynamics. The specific relationship between βmax and We number under different magnetic field intensities (B) shows that a vertical magnetic field has a great inhibiting effect on liquid metal droplet spreading. Finally, the influence of oxidation on droplet spreading characteristics also has been studied. These basic findings are important for the application of liquid metal on a divertor/limiter in a fusion reactor, offering a theoretical reference engineering design.

Conceptual design of a 3D printed liquid lithium divertor test modular for EAST

The utilization of Liquid Metal Divertor (LMD) stands as a promising prospect for the divertor component in DEMO. To explore the vapor shielding effect and provide technical accumulation for the development of DEMO LMD, the imperative development of a high-performance LMD test module has arisen. The Experimental Advanced Superconducting Tokamak (EAST) has amassed substantial expertise in LM first wall design and experimentation, rendering it suitably equipped for experimental pursuits. Therefore, this paper aims to design a test module running in the mid-plane of EAST to accumulate technical experience for the application of 3D printed capillary porous system (CPS) in LMD. Drawing upon prior experience, specific design requirements are articulated. Lithium is chosen as the operational LM due to its favorable balance between heat dissipation capacity and compatibility with the core plasma. The conceptual design encompasses active water cooling components, a CPS, and a lithium channel. Thermal analysis is conducted to ascertain the prevention of excessive evaporation or frozen during operation. The maximum power handling capacity of this design is projected, with practical solutions outlined for inter-component thermal management. The results affirm the viability of this design within the thermal load parameters of the EAST experiment, offering invaluable insights for designers exploring diverse LMD configurations.

Depth profile CF LIBS analysis of the wall deposited layer in the COMPASS tokamak after LiSn testing campaign

Plasma-facing materials in tokamak nuclear reactors are subject to intense heat and radiation. The design of liquid metal (LM) divertor walls has several advantages over the solid counterparts. It is important to understand the migration and re-deposition of LMs such as lithium (Li), lead (Pb) or even tin (Sn). This work presents CF LIBS depth profile analysis of the deposited layer on Ni–Cr based screws situated at different places in the COMPASS tokamak vessel during the Capillary Porous System (CPS) LM (LiSn) divertor testing campaign. The experiment was carried out at low Ar pressure (100 mbar) using ns Nd:YAG laser operating at 532 nm. The spectra were recorded using echelle spectrometer (M5000, Andor Tech.) equipped with an iCCD camera (iStar DH743, Andor Tech.). The quantification was based on observed C I, B I, Li I and (weak) Sn I lines, in addition to trace elements (Ca, Na, K) and main matrix elements (Cr, Ni, Fe). The evaluation of electron temperature and density was based on Cr I-II, Ni-II and Fe I-II lines. The LM based content is estimated to be Li (0.3-0.81 at.%) and Sn (0.1-0.23 at.%) with Li/Sn at.% ratio ranging from 3 to 3.5 (wt.% ratio of 0.17-0.20). In practise, the limit of detection of Sn by CF LIBS for this type of Cr-Ni-Fe based matrix is observed to be ∼0.1 at.%.

Hydrogen interaction with solid and liquid tin

Hydrogen interacts with most metals, and basic thermodynamic parameters enable estimating equilibrium concentration in a wide range of pressures and temperatures. Tin is a common metal whose hydrogen solubility has not been adequately determined yet. A series of solid and liquid tin experiments within the temperature range from 150 to 350 °C exhibit high inertness to hydrogen. Its equilibrium concentration at the melting point of 231.9 °C and 1 bar pressure was below the detection limit of 1 ppb H/Sn. Similarly, the deuterium permeation rate j through the 0.25 mm thick tin membrane at 200 °C was below ∼2.5 × 109 D2/cm2/s, the uncertainty limit due to the isotope exchange reaction forming HD. The main experimental challenges were eliminating weak hydrogen interaction with tin oxide atop the solid and liquid tin and suppressing the setup's hydrogen background outgassing. Tin's inertness to molecular hydrogen may have relevance in some specific technologies. Unfortunately, it has been recently published that tin interacts intensively with hydrogen plasma, which is incompatible with liquid metal divertor demands.

Performance of a liquid Sn divertor target during ASDEX upgrade L-mode and H-mode operation

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 reaching the main plasma, which may have originated from either the CPS or leaked tin. This conclusion is also supported by splashes of tin droplets, which were observed on the adjacent divertor tile and one ∼0.5m downstream. Therefore, to make a Sn-filled LMD a viable alternative to solid tungsten, the formation of droplets must be reduced by two orders of magnitude.

Towards Integrated Target–SOL–Core Plasma Simulations for Fusion Devices with Liquid Metal Targets

Self-healing liquid metal divertors (LMDs) based on the Capillary Porous Structure (CPS) concept are currently being considered among the possible solutions to the power exhaust problem in future fusion reactors. Indeed, the passive replenishment of the plasma-facing surface by capillary forces and the self-shielding of the target via vapor emission can potentially improve the divertor lifetime and its resilience to transient loads. On the other hand, the LMD target erosion can be significant due to evaporation and thermal sputtering, on top of physical sputtering, possibly leading to unacceptable core plasma dilution/power losses (for a low-Z/high-Z metal such as Li and Sn, respectively). For this reason, it is necessary to assess whether an LMD is compatible with an European DEMO (EU-DEMO) plasma scenario. This requires a self-consistent model of the impurity emission from the target, the plasma in both the scrape-off layer (SOL) and the core regions and the transport of impurities therein. In this paper, an an integrated modelling approach is proposed, which is based on SOLPS-ITER and includes its coupling with a target erosion model written in FreeFem++ and a core plasma model (ASTRA/STRAHL). An application of the coupled SOL-target model to simulate experiments performed in the Magnum-PSI linear plasma device with a CPS target filled with Li is also included to provide a first demonstration of the capabilities of the approach. Results are promising, being in good agreement (within a few degrees) with the measured target temperature distribution. In perspective, the modelling framework presented here will be applied to the EU-DEMO with an Sn divertor.

Predictive and interpretative modelling of ASDEX-upgrade liquid metal divertor experiment

A liquid metal Capillary Porous System (CPS) test module filled with tin was studied in the ASDEX Upgrade (AUG) outer divertor. The CPS module was flush mounted as part of a target tile and exposed using the AUG divertor manipulator. In order to predict tin erosion from the designed module under typical AUG divertor loading conditions, the experiment was interpreted using the HeatLMD code. Preceding test exposures of the CPS in the high heat flux facility GLADIS were performed and interpreted by modelling to quantify the thermo-mechanical properties of the module. The results for the reference AUG discharge indicated a total of 2.6 × 1017 tin atoms (51 μg) would be eroded during the exposure, predominantly through temperature enhanced sputtering. The vapour cooling power was predicted to be negligible (5 kW/m2at the end of a 5 s exposure with heat flux from the plasma of 2 MW/m2). The module was expected to be compatible with plasma operation, with tin erosion too low for any significant effect on the plasma performance. However, interpretative modelling of the experimental discharge with the highest exposure time yielded significantly lower tin erosion than observed. To be attributed to tin radiation the experimentally observed increase in total radiative power (1.5 MW) would require 2 × 1018 tin atoms (peak calculated erosion rate) radiating in the core plasma. This would require every tin atom eroded, to reach the core, which is unlikely.

Magnetic control of DTT alternative plasma configurations

One of the main challenges concerning next generation tokamaks (such as DEMO) will be the development of a heat and power exhaust system able to withstand the large loads expected in the divertor region. A dedicated Divertor Tokamak Test (DTT) facility has been proposed in the EUROfusion Roadmap, with the aim of testing unconventional solutions, such as advanced magnetic configurations and liquid metal divertors. Magnetic control of alternative plasma configurations, such as the X-Divertor, will play a key role in the solution of the heat exhaust and yet can be a challenging point, due to increased sensitivity introduced by secondary x-points. To overcome the complications introduced by secondary x-points in advanced plasma shapes, magnetic control in DTT is achieved by resolving to the eXtreme Shape Controller, in order to control both the plasma shape and the secondary x-point position.

Hydrogen diffusion on the tin-covered tungsten surface: A first-principles study

The diffusion behavior of hydrogen (H) on tin-covered tungsten (Sn-W) surfaces is of great significance for understanding the retention of H in Sn-W materials exposed to H plasma. In this work, the diffusion mechanism of H in Sn-W surfaces and the related interactions were studied through employing density functional theory. The results show that Sn atoms are strongly adsorbed on the W surface with energy of -5.10 ∼ -5.60 eV with surface reconstruction for W (100). The diffusion of H atom from top of Sn to interior on W (100) and W (110) surface requires energy barrier of at least 1.02 eV and 0.76 eV, respectively. More interestingly, the Sn converging on W surface can reduce the diffusion energy barriers of H atom on or into the W surface, which indicates that the diffusion of H atom became easier in comparison to naked W surface. The facilitation effect of Sn on H atom diffusion can be attributed to the strong interaction between the Sn atom and W surface as proved by the calculated charge density and interaction energy. However, the coverage of Sn has little impact on the diffusion of H atom on or into the W sub-surface. As the depth increases, the influence of surface Sn atoms on the diffusion barrier of H decreases. The influence of liquid tin on the diffusion of hydrogen in solid tungsten will be helpful to understand the abnormal retention mechanism for liquid metal divertor.

Commissioning and first results of the OLMAT facility

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.

Simulation of Liquid Lithium Divertor Geometry Using SOLPS-ITER

Plasma-facing component (PFC) geometries are evaluated for a Fusion Nuclear Science Facility (FNSF) design to find solutions that are compatible with flowing liquid lithium components while satisfying requirements from the perspective of plasma and neutral particle transport. Flowing liquid metal (LM) divertor systems offer important advantages due to continual regeneration of the surface material, but may require modifications from standard optimized tokamak divertor designs. Scrape Off Layer Plasma Simulator for the International Thermonuclear Experimental Reactor (SOLPS-ITER) simulations are used to compare standard vertical target, open, and balanced baffled geometries. It is found that the open geometry offers little control of neutral particles and has narrow operating windows where the divertor fluxes can be lowered to acceptable levels, while maintaining the required upstream density. The addition of neutral baffles allows greater control over the divertor conditions with less impact on the upstream density. The divertor neutral pressure in the baffled geometry is slightly higher than the open divertor, but not as large with a vertical target. A first-pass coupling of the SOLPS solution with sheath and implantation/erosion models to compute the lithium emission shows that the lithium is largely confined to the near-surface region due to the small ionization mean free path and large main ion flow toward the surface. The lithium source was varied by three orders of magnitude. At the highest rate, when the sourced lithium approaches the level of the main ion recycling flux, a high core lithium concentration (~10%) is observed with only moderate power dissipation to affect the divertor plasma.

Divertor geometry modeling with the SOLPS-ITER code for reactor concepts with liquid metal divertors

• The SOLPS-ITER code has been applied to the analysis of the scrape-off layer (SOL) plasma associated with the fast flow Lithium (Li) divertor design for Fusion Nuclear Science Facility (FNSF). • Two flat divertor configurations are investigated and we found that simplified open geometry presents challenges for neutral control and impurity retention, but balanced geometry with baffling appears promising. • Lithium (Li) is sourced into the balanced configuration and scanning over a large range to understand what level of Li emission is needed to influence the divertor and upstream plasma conditions. • At low sourcing level ( ϕ Li ≤ 1 × 10 23 / s ), the Li shows a minor effect on the plasma and is well confined in the Private flux region (PFR) and divertor surfaces. • At high Li sourcing level ( ϕ Li > 1 × 10 24 / s ), the Li strongly affects the upstream and divertor plasma conditions and leads to strong detachment by increasing momentum and power losses. The SOLPS-ITER code has been applied to the analysis of the scrape-off layer (SOL) plasma associated with the fast flow Lithium (Li) divertor design for Fusion Nuclear Science Facility (FNSF). Two divertor configurations are investigated to determine puff levels that provide robust operational windows that meet the FNSF design requirements on upstream density and divertor flux with a flat divertor for flowing liquid metal (LM). Neon (Ne) is introduced to control the heat flux to the divertors while Deuterium (D 2 ) puffing is performed to maintain sufficiently high upstream density. We found that a simplified open geometry presents challenges for neutral control and impurity retention, but a balanced geometry with baffling is found to have acceptable puff operational windows. Lithium (Li) is sourced into the balanced configuration uniformly along the divertor surfaces and scanning over a large range to understand what level of Li emission is needed to influence the divertor and upstream plasma conditions. At low sourcing level ( ϕ Li ≤ 1 × 10 23 / s ), the Li shows a minor effect on the plasma and is well confined in the private flux region (PFR) and divertor surfaces. At moderate sourcing level ( 1 × 10 23 < ϕ Li ≤ 1 × 10 24 / s ), the Li starts affecting the upstream and divertor plasmas, and Li radiation and electron cooling are almost linearly increasing with the Li sourcing level. At high Li sourcing level ( ϕ Li > 1 × 10 24 / s ), the Li strongly affects the upstream and divertor plasma conditions and leads to strong detachment by increasing momentum and power losses. At high sourcing level, Li reaches above the x-point that can result in a substantial Li radiation in the core and sufficient dissipation to strongly affect the divertor and upstream plasmas.

SOLPS-ITER simulations of a CPS-based liquid metal divertor for the EU DEMO: Li vs Sn

In this work, we study the effect of installing a liquid metal divertor (LMD) using a capillary-porous structure in the EU DEMO tokamak within the same envelope of the baseline solid divertor. We used the SOLPS-ITER code to model the scrape-off layer (SOL) plasma and neutrals, coupled to a target thermal model to enable the self-consistent calculation of the LM target erosion rate, and adopting a fluid neutral model for the sake of simplicity. First calculations considering only D and Li (or Sn) showed a significant reduction of the steady state target heat load with respect to simulations considering only D, thanks to vapor shielding. Nevertheless, the computed peak target heat flux (∼31 MW m−2 and ∼44 MW m−2 for Li and Sn, respectively) was still larger than/borderline to the power handling limit of the LMD concepts considered. Moreover, the impurity concentration in the pedestal—a proxy for the core plasma dilution/contamination—was computed to be above/close to tolerability limits suggested by previous COREDIV calculations. These results indicate that the operational window of an LMD for the EU DEMO, without any additional impurity seeding, might be too narrow, if it exists, and that Sn looks more promising than Li. A second set of calculations was then performed simulating Ar seeding in the SOL, to further reduce the target heat load, and consequently the metal erosion rate. It was found that the mitigation of the plasma heat load due to Ar radiation in the SOL effectively replaces the radiation associated to vapor shielding in front of the target, thus allowing to operate the LMD in a regime of low target erosion. The resulting operational window was found to be significantly wider, both in terms of tolerable peak target heat flux and of acceptable core plasma contamination.

Identification of the Postulated Initiating Events of Accidents of a CPS-Based Liquid Metal Divertor for the EU DEMO Fusion Reactor

A preliminary but systematic safety analysis of a liquid metal divertor (LMD) for the EU DEMO performed by means of the Functional Failure Mode and Effect Analysis (FFMEA) is presented. This methodology is suitable for the analysis of the LMD, which is undergoing preconceptual design. In fact, the FFMEA compensates for the lack of detailed design information by postulating the loss of a system function, rather than a specific component failure. The implementation of the FFMEA led to a better understanding of the safety and operational issues associated with the system and to the identification of a list of postulated initiating events (PIEs), i.e., the most challenging conditions for the plant. The PIEs, together with their possible consequences, represent an input for future quantitative safety analyses. Due to the early design stage of the LMD and the iterative nature of the methodology, this list will evolve alongside the design detail and with improvements in the understanding of phenomena driving reactor behavior. The study highlighted some safety-relevant issues, e.g., those related to materials compatibility and system modularity, to be addressed in the perspective of a safety-driven design evolution.

Calibration-free laser-based spectroscopic study of Sn-based alloys

The elemental quantification of liquid metal divertor (LMD) surface is important for understanding the material erosion, migration, re-deposition, and fuel retention in Plasma-Facing Components (PFCs). Currently, LMD are attractive candidates for the short- and long-term operation of fusion devices like DEMO. Liquid metals can provide self-cooling, self-replenishing plasma-facing surfaces requiring very little upkeep. In a previous work, we studied Li and LiSn layers deposited on attachment screws in the COMPASS tokamak by means of Calibration-Free Laser-Induced Breakdown Spectroscopy (CF-LIBS). Several problems were encountered related to the detection of Sn in LiSn. Thus, in the present work, we are optimizing the experimental conditions for the detection of Sn I-II and Pb I-II in Pb-containing Sn-based alloys, performing the quantification of Pb in traces and in bulk quantities using CF-LIBS approach.

Predictive modelling of liquid metal divertor: from COMPASS tokamak towards Upgrade

Following ELMy H-mode experiments with liquid metal divertor target on the COMPASS tokamak, we predict the behavior of a similar target on COMPASS Upgrade, where it will be exposed to surface heat fluxes even higher than those expected in the future EU DEMO attached divertor. We simulate the heat conduction, sputtering, evaporation, excitation and radiation of lithium and tin in the divertor area. Measured high-resolution data from COMPASS tokamak were rescaled towards the Upgrade based on many established scalings. Our simulation then yields the amount of released metal which ranges from 4 mg s−1 upto 12 g s−1 depending mainly on the geometry and Li/Sn choice, quite independently from active cooling or strike point sweeping.

Requirements and interface management for the divertor system design and integration in the Divertor Tokamak Test (DTT) facility

The construction of a Divertor Tokamak Test (DTT) facility is fully included in the European effort towards the realization of fusion energy with the aim of studying possible alternative solutions, with respect to the ITER research program, to tackle the problem of power exhaust. This means that the machine must be compatible with different possibilities for the divertor cassette design and the plasma scenario. The design of first “reference” divertor is largely based on the ITER-like divertor, but in the future other solutions are going to be tested, including liquid metal divertors. To provide a justification basis for the design of the reference cassette and guidelines for the future experimentations it is fundamental to define the requirements of this power exhaust system, in particular the heat flux compatibility and the interfaces with the surrounding interacting systems of the tokamak (the water cooling pipes, the divertor magnetic coils, the divertor diagnostics, etc.). This work shows how the divertor system can be included in a wider Requirement Management procedure that has been defined for DTT highlighting the relevant interface parameters. Additionally, a data analysis tool has been developed to provide an interactive visualization of requirements traceability to directly show the impact of changes in the divertor on the design and operations of the machine and viceversa. In this way, during future studies on alternative design divertors or plasma scenarios, the limiting constraints on both sides can be easily individuated and traced to their motivation, and possibly discussed.

Conceptual design of a liquid-metal divertor for the European DEMO

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 potentially major improvement.