Steady-state Heat Load Research Articles

Conceptual design of coolant circuits and thermal stress analysis for JA-DEMO divertor

Engineering design of the JA-DEMO divertor concept, i.e. double-coolant circuits of 200°C coolant (5MPa) for high heat load targets (CuCrZr-pipe) and 290°C coolant (15MPa) for high neutron load Plasma Facing Units (F82H-pipe) and cassette body (CB), has been developed. Computational fluid dynamics (CFD) calculation of the coolant distributions to targets, baffles, reflectors, dome and to CB was performed in order to determine the feasibility of the key concepts. (i) All PFU support designs for the coolant distribution to PFUs were optimized to reduce variation of the flow velocity (Vcool) at the inlet of PFUs less than 5 – 6 %. Parallel coolant circuit design for the dome and reflectors was also developed, and mass flows were adjusted by orifices and the main mass flow rate. (ii) A new cooling concept with two layers of puddles and fins at the side routes was proposed for CB. The design provided Vcool = 0.55 – 1.27 m⋅s−1 and average Vcool = 1.04 m⋅s−1 with the fin transparent ratio of 0.1. It was enough to exhaust the nuclear heat, and the design issues were identified under the coolant condition. (iii) Heat exhaust on fish-scale surface target and stress-strain of the CuCrZr pipe were investigated under assuming a DEMO divertor condition. Steady-state heat load at wet region by plasma (qtwet) was restricted below 13.5 MWm−2 to avoid W-recrystallization. In the stress-strain cycle of higher qtwet ∼15.3 MWm−2, maximum tensile σ (120 – 200 MPa) and total change in strain during the heat load cycle (Δε ∼0.25 %) were relatively small. These evaluations suggested that such high heat load cycle may not be a critical lifetime issue, but reduction in Tcool is preferable to handle ITER-like slow transients such as 15 – 20 MWm−2.

Design of a 3D-printed liquid lithium divertor target plate and its interaction with high-density plasma

A liquid Li divertor is a promising alternative for future fusion devices. In this work a new divertor model is proposed, which is processed by 3D-printing technology to accurately control the size of the internal capillary structure. At a steady-state heat load of 10 MW m−2, the thermal stress of the tungsten target is within the bearing range of tungsten by finite-element simulation. In order to evaluate the wicking ability of the capillary structure, the wicking process at 600 °C was simulated by FLUENT. The result was identical to that of the corresponding experiments. Within 1 s, liquid lithium was wicked to the target surface by the capillary structure of the target and quickly spread on the target surface. During the wicking process, the average wicking mass rate of lithium should reach 0.062 g s−1, which could even supplement the evaporation requirement of liquid lithium under an environment > 950 °C. Irradiation experiments under different plasma discharge currents were carried out in a linear plasma device (SCU-PSI), and the evolution of the vapor cloud during plasma irradiation was analyzed. It was found that the target temperature tends to plateau despite the gradually increased input current, indicating that the vapor shielding effect is gradually enhanced. The irradiation experiment also confirmed that the 3D-printed tungsten structure has better heat consumption performance than a tungsten mesh structure or multichannel structure. These results reveal the application potential and feasibility of a 3D-printed porous capillary structure in plasma-facing components and provide a reference for further liquid−solid combined target designs.

Manufacturing trials of PFCs with low thermal conductivity features for limiters

In future demonstration fusion power plants, plasma facing components will face high steady state and ultra-high transient heat loads from the plasma. Introducing low thermal conductivity features to the component can reduce the peak temperatures seen by the coolant pipe during transient loads, but at the trade-off of a loss of steady-state performance. Producing low-conductivity tungsten by additive methods has been investigated elsewhere, so this trial investigates production through subtractive means i.e., conventional milling. This method preserves the original temperature limits and majority of the microstructure and is much higher Technology Readiness Level (TRL) so requires less development to reach series manufacture. In this trial, diamond drilling produced holes in monoblocks down to 0.6 mm and ligaments down to 0.7 mm, showing that this method is capable of producing small features in a controlled way. With the geometric design of lattice used in this trial, these samples achieved thermal conductivity reductions of 15% to 40%. One monoblock assembly was tested up to 1100 °C for 500 cycles in the Heating by Induction to Verify Extremes (HIVE) High Heat Flux test rig and showed no signs of ligament cracking or thermal degradation. Electromagnetic modelling has been validated against the experimental results, so future designs can be accurately modelled ahead of testing in HIVE.

Effects of Y2Zr2O7 particles and rolling deformation process on the mechanical properties and heat load behavior of W–Y2Zr2O7 plate

This paper addresses the significant requirements in China's fusion engineering and investigates the preparation technology and performance characterization of high-performance tungsten-based materials used as PFMs. The rolled W–Y2Zr2O7 plate were prepared through sintering and subsequent rolling, and steady-state heat load assessment was conducted under fusion device working conditions. Additionally, the change in mechanical behavior of the rolled plates was investigated across different surfaces. The results also revealed that the rolling plate exhibited a DBTT between RT and 200 °C, with more superior mechanical properties along the RD direction. The flat-plate W–Y2Zr2O7/Cu components demonstrated the ability to withstand 1000 heat load cycles at 10 MW/m2 without cracking, indicating favorable thermo-mechanical performance as PFMs. These findings serve as a valuable reference for the fabrication of tungsten-based materials in fusion engineering.

Long-term plasma exposure of ITER-like W/Cu monoblocks with pre-damaged surfaces in EAST experiments

In the ITER and future fusion devices, W/Cu monoblocks will be used as divertor target which are exposed to both steady state heat load and transient heat flux. Especially, the transient heat flux up to 10 GW m−2 during plasma disruption, is expected to induce the shallow surface damages, such as melting, and even boiling of W/Cu monoblocks. Thus, the performance of damaged W/Cu monoblocks under subsequent long-term plasma discharges is a key concern that needs to be verified and tested on existing tokamaks. Since 2022, a new type of main limiter composed of ITER-like W/Cu monoblocks has been installed and tested in EAST. The surface of W/Cu monoblocks of the limiter was damaged by the transient heat flux during the early stages of plasma construction. Subsequently, they were subjected to long-term plasma discharges over 2600 shots in normal plasma discharge conditions. This circumstance conveniently facilitates the discussion of the performance of W/Cu monoblocks with damaged surfaces especially a melting edge with hill structure under prolonged exposure to plasma. In general, the shallow damage resulting from transient heat flux on W/Cu monoblocks appears to have minimal impact on the heat exhaust capacity under steady-state heat loads, as indicated by both experimental monitoring and numerical simulation results. However, shallow melting, leading to a change in surface structure and the formation of hills, could theoretically increase local temperatures, creating potential hot spots. This phenomenon requires further validation through dedicated experiments. Moreover, the brittleness of the near-surface layer may give rise to brittle destructions, such as cracks and even dust particles, posing an additional concern. These findings yield unique qualitative conclusions that can be referenced for ITER and other fusion devices.

Electron beam profile measurement using enhanced dual-techniques in high heat flux test facility at Institute for Plasma Research.

During operation of an ITER-like plasma fusion device, Plasma Facing Components (PFCs) of a divertor are subjected to steady-state and transient heat loads of up to 20 MW/m2. At High Heat Flux Test Facility (HHFTF) at Institute for Plasma Research, PFCs are subjected to such heat flux scenarios using 200 kW electron beam (EB). A heat flux distribution on PFCs is significantly influenced by the EB profile and scanning technique. A novel dual-technique approach is used to measure EB cross-sectional profiles in HHFTF. A single setup is designed, fabricated, and installed in HHFTF consisting of (1) the knife-edge technique involving graphite and tungsten rectangular tiles bonded on a copper block adjacent to each other to form a sharp edge and (2) the wire scanner technique wherein tungsten wires are placed at equal distances. Current collected by the setup is measured as EB is rastered across the wire and tile boundaries. Current measured as a function of EB position is used to calculate the EB profile using a customized LabVIEW software application. Experiments are conducted by varying EB power from 28 to 200 kW and correspondingly the full width at half maximum of the EB profile grows from 5.30 to 16.04mm, as observed by both these techniques. The COMSOL Multiphysics simulation of the wire scanner technique is performed to verify the experimental results. X-ray and infra-red imaging diagnostics of HHFTF are also used to measure the EB profile in a unique and first-of-its-kind way, and they are found to agree well with the dual-technique measurements.

Prospects of negative triangularity tokamak for advanced steady-state confinement of fusion plasmas in MHD stability consideration

The steady-state confinement, beta limit, and divertor heat load are among the most concerned issues for toroidal confinement of fusion plasmas. In this work, we show that the negative triangularity tokamak has promising prospects to address these issues. We first demonstrate that the negative triangularity tokamak generates the filed line rotation transform more effectively. This brings bright prospects for the advanced steady-state tokamak scenario. Given this, the MHD stability and equilibrium confinement of negative triangularity tokamak are investigated. We point out that the negative triangularity configuration with a broad pressure profile is indeed more unstable for low-n magnetohydrodynamic modes than the positive triangularity case so that the H-mode confinement can hardly be achieved in this configuration, where n is the toroidal mode number. Nevertheless, we found that the negative triangularity configuration with high bootstrap current fraction, high poloidal beta, and peaked pressure profiles can achieve higher normalized beta for low-n modes than the positive triangularity case. In a certain parameter domain, the normalized beta can reach about twice the extended Troyon limit, while the same computation indicates that the positive triangularity configuration is indeed constrained by the Troyon limit. This shows that the negative triangularity tokamaks are not only favorable for divertor design to avoid the edge localized modes but also can have promising prospects for advanced steady-state confinement of fusion plasmas in high beta.

Neutron irradiation effects on the low cycle thermal fatigue performance of first wall mock-ups

For the first wall (FW) of ITER, beryllium will be the plasma facing material and has to sustain high thermal, particle and neutron loads. In order to assess the synergistic effects of thermal and neutron loads in total ten Be-armored FW normal heat flux (NHF) mock-ups were produced consisting of two Be flat tiles each joined via hot isostatic pressing (HIP) to a CuCrZr heat sink and a steel support structure by different manufacturing routes to assess the most promising ones. Five of them were neutron irradiated, two at Centrum Výzkumu Řež in Czech Republic and three in the High Flux Reactor at NRG in the Netherlands. The remaining mock-ups were kept as reference. All flat-tile mock-ups had two Be-tiles and were exposed to cyclic steady state heat loads in the electron beam facilities JUDITH 1 and 2 up to a maximum power density of 3.75 MW/m2, in order to find the damaging threshold. A screening step using 1 MW/m2 was performed after finishing each loading step and after failure for direct comparison and detection of any deterioration caused by the cycling. Despite being still in the early development phase of Be joining, all mock-ups sustained the cycling up to at least 2.75 MW/m2 and clear differences in the performance of irradiated vs. non-irradiated mock-ups were observed.

Thermohydraulic and thermal fatigue characteristics of W/Cu flat-type microchannel mock-up for CFETR divertor

The divertor of the China Fusion Engineering Test Reactor (CFETR) requires a more efficient heat dissipation structure due to the high demand of heat flux dissipation of up to 20 MW/m2. Microchannel cooling technology excels in thermal management because of large heat transfer area to volume ratio and capability to achieve uniform temperature distribution. However, the heat transfer and thermal fatigue characteristics of W/Cu flat type mock-up with microchannel cooling structures at a heat flux of 20 MW/m2 remains elusive. In this work, we analyzed the thermohydraulic (numerically) and thermal fatigue (experimentally) characteristics of W/Cu flat-type microchannel mock-up. The numerical results show that the thermal loaded area (W) of the microchannel mock-up possesses excellent temperature uniformity and heat transfer performance (the W surface temperature is only ∼685 °C even at a low volumetric flow rate V̇ = 5 L/min). We also compared the heat dissipation efficiency with the hypervapotron cooling channels, and found that the maximum temperature of the W surface and channel wall of the microchannel mock-up are reduced by 18.9 % and 46.7 %, respectively. The thermal stress analysis indicates that the maximum equivalent stress is concentrated in the middle region of the W surface at 15 s, with a value of 357.8 MPa that is below the maximum allowable stress of W material (425 MPa at 650 °C). The as-manufactured W/Cu flat-type microchannel mock-up successfully withstood a thermal fatigue test consisting of 1000 cycles at 20 MW/m2 steady state heat load at V̇ = 30 L/min and Tin = 25 °C. The maximum temperature of W is only 616.2 °C during the test, and some microcracks are detected on the W surface after the test; these microcracks are perpendicular to the heat loading surface, which have negligible influence to the heat transfer of the entire mock-up. This study reveals that the W/Cu flat-type microchannel mock-up is promising in dissipating ultra-high heat flux (steady-state 20 MW/m2) on the divertor target, which might provide valuable insights for the engineering design of the CFETR divertor.

Testing of ITER-grade plasma facing units in the WEST tokamak: Progress in understanding heat loading and damage mechanisms

Assessing the performance of the ITER design for the tungsten (W) divertor Plasma Facing Units (PFUs) in a tokamak environment is a high priority issue to ensure efficient plasma operation. This paper reviews the most recent results derived from experiments and post-mortem analysis of the ITER-grade PFUs exposed in the WEST tokamak and the associated modelling, with a focus on understanding heat loading and damage evolution. Several shaping options, sharp or chamfered leading edge (LE), unshaped or shaped blocks with a toroidal bevel as foreseen in ITER, were investigated, under steady state heat fluxes of up to 120 MW⋅m−2 and 6 MW⋅m−2 on the sharp LE and top surface of the block, respectively. A very high spatial resolution (VHR) infrared (IR) camera (0.1 mm/pixel) was used to derive the temporal and surface distribution of the temperature and heat load on the castellated tungsten blocks for different geometric alignment and plasma conditions. Photonic modelling was required to reproduce the IR measurements in particular in the toroidal and poloidal gaps of the mono-block (MB) stacks where high apparent temperatures are observed. Specular reflection is found to be the dominant emitter in these parts of the blocks. W-cracking was observed on the leading edge of the blocks already within the first phase of plasma operation, during which the divertor was equipped with unshaped PFUs, including some intentionally misaligned blocks. Numerical analysis taking into account softening processes and mechanical stresses, revealed brittle failure due to transients as the dominant failure mechanisms. Ductile failure was observed in one particular block used for the melting experiment, therefore under extremely high steady state heat load conditions. W-melting achieved on actively cooled PFU exhibits specific features: shallow melting and slow melt displacement. Plasma exposure of pre-damaged PFUs at various damage levels (crack network and melted droplets) was carried out under high heat load conditions with several hours of cumulated plasma duration. IR data and preliminary surface analyses show no evidence of significant degradation damage progression under these conditions.

Tungsten based divertor development for Wendelstein 7-X

Wendelstein 7-X, the world’s largest superconducting stellarator in Greifswald (Germany), started plasma experiments with a water-cooled plasma-facing wall in 2022, allowing for long pulse operation. In parallel, a project was launched in 2021 to develop a W based divertor, replacing the current CFC divertor, to demonstrate plasma performance of a stellarator with a reactor relevant plasma facing materials with low tritium retention. The project consists of two tasks: Based on experience from the previous experimental campaigns and improved physics modelling, the geometry of the plasma-facing surface of the divertor and baffles is optimized to prevent overloads and to improve exhaust. In parallel, the manufacturing technology for a W based target module is qualified.This paper gives a status update of project. It focusses on the conceptual design of a W based target module, the manufacturing technology and its qualification, which is conducted in the framework of the EUROfusion funded WPDIV program. A flat tile design in which a target module is made of a single target element is pursued. The technology must allow for moderate curvatures of the plasma-facing surface to follow the magnetic field lines. The target element is designed for steady state heat loads of 10 MW/m2 (as for the CFC divertor). Target modules of a similar size and weight as for the CFC divertor are assumed (approx. < 0.25 m2 and < 60 kg) using the existing water cooling infrastructure providing 5 l/s and roughly maximum 15 bar pressure drop per module.The main technology under qualification is based on a CuCrZr heat sink made either by additive manufacturing using laser powder bed fusion (LPBF) or by uniaxial diffusion welding of pre-machined forged CuCrZr plates. After heat treatment, the plasma-facing side of the heat sink is covered by W or if feasible by the more ductile WNiFe, preferably by coating or alternatively by hot isostatic pressing W based tiles with a soft OFE-Cu interlayer. Last step is a final machining of the plasma-exposed surface and the interfaces to the water supply lines and supports to correct manufacturing deformations.

Development of pebble-based extruded carbon rods for extreme plasma heat flux environments

This work presents first experiments toward the development of continuously renewable (extrudable) pebble-based carbon rods for use as plasma-facing components in extreme steady-state plasma flux environments. The primary envisioned application of this work is a first wall that can survive long-term in future magnetic fusion power reactors while also improving recovery of the reactor fuel (tritium and deuterium atoms). Bench tests applying extreme steady-state front-surface heat loads of up to 50 MW/m2 are presented. Continuous pebble rod front-surface recession and intact pebble recovery are successfully demonstrated, at a rate of order 0.2 cm/s. Numerical simulations of the pebble rod front-surface recession are able to match observations reasonably well, indicating that the recession mechanism can be understood as occurring due to pebble thermal expansion and resulting shock and cracking of the inter-pebble binder. Tests of the pebble rod extrusion demonstrate that friction between the rods and the stainless steel extrusion channel is tolerably low (&amp;lt;50 N for the expected channel length) over a wide range of temperatures. Front-surface outgassing rates below 1000 Torr L/s/m2 are achieved, believed to be sufficiently low for use in magnetic fusion reactors. Initial parametric scans over pebble rod size and binder fraction to vary front-surface recession rates are presented.

Long-pulse H-mode operation with stored-energy monitoring for detachment feedback control with a new lower tungsten divertor in EAST

One of the key challenges for future fusion research is to mitigate the high steady-state heat load on the divertor target plates, and divertor detachment offers a promising solution. The Experimental Advanced Superconducting Tokamak (EAST) has developed several feedback control methods for divertor detachment. However, when an off-normal event momentarily disturbs the main plasma, impurity seeding may still be conducted by these methods for detachment, which probably drives the main plasma further away from its stable equilibrium or even causes the disruption. These off-normal events include excessive impurity seeding, loss of heating and dust droplets, which are not rare in current tokamak experiments, especially in long-pulse operation. To compensate for the drawbacks of these methods, we propose and develop a module of stored-energy monitoring to ensure stable plasmas in long-pulse operation. The stored energy usually decreases when the main plasma is away from its stable equilibrium, which is suitable for monitoring the state of the main plasma. Once the stored energy falls below a certain threshold, the module actively switches off the impurity seeding system. Without impurity seeding, the main plasma can recover with the increase in the stored energy. Only when the stored energy exceeds another threshold does the module switch on the impurity seeding to continue the detachment operation. The module function has been verified during the EAST radiative divertor experiments in the newly-upgraded lower tungsten divertor. A typical ∼20 s discharge in grassy-ELM H-mode regime with ∼5 MW source heating power is demonstrated with divertor partial detachment and good energy confinement by active impurity seeding (50% neon, 50% D2). The energy confinement factor is maintained at a high level, i.e. The electron temperature in the core region only has a slight change after the impurity seeding, while the electron density has a ∼10% increase. Furthermore, the ion temperature near the axis also has a remarkable increase. These achievements provide an important demonstration of the actively controlled radiative divertor mitigating the heat loads with good core confinement, which is an essential step toward steady-state operation of fusion reactors.

Conceptual design of Fiber Bragg Grating temperature sensors for heat load measurements in COMPASS-U plasma-facing components

Information about the temperature of plasma-facing components is important for a reliable tokamak operation. A temperature monitoring system using Fiber Bragg Grating (FBG) sensors is foreseen for the new tokamak COMPASS Upgrade, which is currently starting its construction. This diagnostic can be a valuable complement to IR thermography, thermocouples, and Langmuir probe divertor diagnostics. The system will be optimized to estimate the steady-state and transient heat loads, such as runaway electrons, on the divertor and limiters. In this contribution, current progress in the design of the FBG sensors for the COMPASS-U initial open divertor and guard limiters is presented. The heat flux on the plasma-facing components is modeled for diverted and first phase circular plasma scenarios by the PFCFlux code. The subsequent heating and the mechanical strain in the components dedicated for the placement of the sensors is simulated by the finite element analysis software ANSYS. Using these results, the optimal placement and configuration of the sensors are determined with respect to the anticipated maximum temperature and gradients.

Novel SOLPS-ITER simulations of X-point target and snowflake divertors

The design and understanding of alternative divertor configurations may be crucial for achieving acceptable steady-state heat and particle material loads for magnetic confinement fusion reactors. Multiple X-point alternative divertor geometries such as snowflakes and X-point targets have great potential in reducing power loads, but have not yet been simulated widely in codes with kinetic neutrals. This paper discusses recent changes made to the SOLPS-ITER code to allow for the simulation of X-point target and low-field side snowflake divertor geometries. Snowflake simulations using this method are presented, in addition to the first SOLPS-ITER simulation of the X-point target. Analysis of these results show reasonable consistency with the simple modelling and theoretical predictions, supporting the validity of the methodology implemented.

Preliminary exploration of a WTaVTiCr high-entropy alloy as a plasma-facing material

With great power comes great challenges. For nuclear fusion, the holy grail of energy, taming the flame of a miniature star in a solid container remains one of the most fundamental challenges. A tungsten armour for the solid container marks a temporary triumph—a solution adopted by the world’s largest fusion experiment, ITER—but may be insufficient for future challenges. High-entropy alloys (HEAs), which are characteristic of a massive compositional space, may bring new solutions. Here, we explore their potential as plasma-facing materials (PFMs) with a prototype W57Ta21V11Ti8Cr3 HEA that was designed by exploiting the natural-mixing tendency among low-activation refractory elements. Revealed by x-ray diffraction analysis and energy-dispersive x-ray spectroscopy, it predominantly consists of a single bcc-phase but with V, Ti, and Cr segregation to grain boundaries and at precipitates. Its yield strength improves ∼60% at room temperature and oxidation rate reduces ∼6 times at 1273 K, compared with conventionally used W. The Ti–V–Cr rich segregations and the formed CrTaO4 compound contribute to the improved oxidation resistance. However, the Ti–V–Cr rich segregations, along with the decreasing valence-electron concentration of the matrix by the addition of Ta, V and Ti elements, considerably increase the deuterium retention of the W57Ta21V11Ti8Cr3 HEA to ∼675 multiples of recrystallized W. Moreover, its thermal conductivity decreases, being ∼40% of W at 973 K. However, the maximum tolerable steady-state heat load is still ∼84% of W because of its exceedingly high yield strength at elevated temperatures. Overall, despite being preliminary, we expect HEAs to play an important role in the development of advanced PFMs, for their disadvantages are likely to be compensated by their advantages or be overcome by composition optimization.

Fishtail Divertor and its basis surface thermal load bearing performance

Fishtail Divertor (FTD) configuration is a dynamic magnetic field divertor configuration that uses the AC coil installed on the back of the divertor target plate to realize the swing of the strike line. The most important driving force of FTD configuration design is to improve the thermal bearing performance of divertor target plate. Under FTD configuration, the maximum temperature of divertor target plate will eventually reach a dynamic equilibrium state, and it have an obvious jump at both ends of the swing area. With the increase of the swing frequency, FTD configuration can more effectively improve the thermal load bearing performance of the divertor target plate. However, if only to alleviate the steady-state heat load, the swing frequency of FTD no need to be higher than 100 Hz. FTD configuration has engineering application prospects in future fusion reactors. Under a swing distance of 10 cm and a thermal load condition of qp=100MW/m2 and λT=5mm (steady state), Triangular wave Fishtail Divertor (TFTD) swing frequency exceeds 5 Hz or Sine wave Fishtail Divertor (SFTD) swing frequency exceeds 20 Hz, the maximum temperature of target plate can be lower than 900 K. At the same conditions expect the swing frequency is 100 Hz, the maximum temperature under TFTD and SFTD can lower than 700 K and 800 K respectively.

Surface Damage and Microstructure Evolution of Yttria Particle-Reinforced Tungsten Plate during Transient Laser Thermal Shock

Tungsten and its alloys are considered to be the most nominated plasma-facing materials in fusion reactors, which will be exposed to enormously rigorous conditions such as thermal load, plasma exposure, and neutron radiation. At present, the research on the behavior of oxide particle-reinforced tungsten-based materials under long-term steady-state heat load and transient thermal shock is insufficient. The purpose of this study is to investigate the performance of yttria particle-reinforced tungsten plates prepared by the wet chemical method under heat loads by means of indirect coupling experiments. An Nd:YAG laser device is used to perform thermal shock events. The surface damage and microstructure evolution of rolled and fully recrystallized samples exposed to laser thermal shock are observed and analyzed. The cracking threshold of the rolled and fully recrystallized samples is about 0.40~0.48 GW/m2; the degree of surface damage of them aggravates with the increased laser power density. What is more, cracks or even melting damage could be observed on the surface and be accelerated by the process of recrystallization, resulting in the degradation of the ability to withstand the thermal shock of the material.

Cross-code comparison of the edge codes SOLPS-ITER, SOLEDGE2D and UEDGE in modelling a low-power scenario in the DTT

As reactor-level nuclear fusion experiments are approaching, a solution to the power exhaust issue in future fusion reactors is still missing. The maximum steady-state heat load that can be exhausted by the present technology is around 10 MW m−2. Different promising strategies aiming at successfully managing the power exhaust in reactor-relevant conditions such that the limit is not exceeded are under investigation, and will be tested in the Divertor Tokamak Test (DTT) experiment. Meanwhile, the design of tokamaks beyond the DTT, e.g. EU-DEMO/ARC, is progressing at a high pace. A strategy to work around the present lack of reactor-relevant data consists of exploiting modelling to reduce the uncertainty in the extrapolation in the design phase. Different simulation tools, with their own capabilities and limitations, can be employed for this purpose. In this work, we compare SOLPS-ITER, SOLEDGE2D and UEDGE, three state-of-the-art edge codes heavily used in power exhaust studies, in modelling the same DTT low-power, pure-deuterium, narrow heat-flux-width scenario. This simplified, although still reactor-relevant, testbed eases the cross-comparison and the interpretation of the code predictions, to identify areas where results differ and develop understanding of the underlying causes. Under the conditions investigated, the codes show encouraging agreement in terms of key parameters at both targets, including peak parallel heat flux (1%–45%), ion temperature (2%–19%), and inner target plasma density (1%–23%) when run with similar input. However, strong disagreement is observed for the remaining quantities, from 30% at outer mid-plane up to a factor 4–5 at the targets. The results primarily reflect limitations of the codes: the SOLPS-ITER plasma mesh not reaching the first wall, SOLEDGE2D not including ion-neutral temperature equilibration, and UEDGE enforcing a common ion-neutral temperature. Potential improvements that could help enhance the accuracy of the code models for future applications are also discussed.

Recent progress of thick tungsten coating prepared by chemical vapor deposition as the plasma-facing material

Chemical vapor deposition (CVD) is a promising technique for the preparation of W-based plasma-facing materials (PFMs). An overview of the microstructure, chemical composition, thermal conductivity, thermal stability, thermal shock performance under disruption-like and edge localized mode-like transient heat load, and neutron irradiation performance of CVD-W has been given in our previous work. However, for fusion applications, additional properties need to be assessed. To this end, deuterium (D) permeability, D plasma irradiation performance, and thermal fatigue resistance of CVD-W were investigated in this work. The results showed that the D permeability of CVD-W in the temperature range of 973–1173 K was larger than that of the commercial pure W, which was related to the columnar grain structure of CVD-W. Additionally, both CVD-W and commercial pure W were exposed to D plasma up to a fluence of 1 × 1026 m−2. Compared to commercial pure W, CVD-W exhibited a mitigated blistering behavior and lower D total retention, which could be attributed to its strong [001] crystallographic texture along the thickness direction and a lower number of defect density (e.g. grain boundaries). CVD-W and commercial pure W were also exposed to steady-state and transient heat load simultaneously, leading to a base surface temperature and surface temperature increase of about 953–1473 K and 250–300 K, respectively. A strong grain orientation dependence of the surface degradation induced by the combined heat load has been found. Consequently, CVD-W exhibited a much more uniform plastic deformation than pure W, and no surface cracks along grain boundaries were observed in CVD-W. Finally, the industrial-scale production of CVD-W-based PFMs and mockups was demonstrated. This work paves the way for the fusion applications of thick CVD-W coatings.