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• Some silicide materials are found out as insert materials for the fabrication of W/CFC blocks. WSi 2 do not exceed but are the same level as insert material as that of SiC + additives. • Zr and V behave as insert materials for W/CFC bonding, but the thermal stability of specimens is not enough. • The bonding strength near edge of an expected size of mono-block is not enough. It needs to be fabricated as a larger size block with 40 MPa of pressure and removed its near-edge parts. The construction of a satellite tokamak, JT-60SA, was completed in 2020. JT-60SA is expected to contribute the research of ITER and developments for DEMO. JT-60SA employs carbon and carbon fiber strengthened carbon (CFC) composites divertors. The divertors will be set on an actively cooled cassette and be remotely handled because of human access restriction into the vacuum vessel due to the radioactivation caused by repeated deuterium plasma operations. An important mission of JT-60SA is high-β steady state plasma research, and during the mission, the use of carbon divertors will be continued. The possibility of metal wall experiments after the complete of high-β steady state plasma research is under discussion. Then the use of remote handled divertor cassette will be a condition, therefore, high performance and appropriate weight divertors are convenient for the operation. One idea is to use clad plates of tungsten and CFC. In previous researches revealed that a sinter bonding method using an insert material of SiC powders + additives was able to join W and CFC. But the research accidentally found out the insert material without explore of the other materials and the size of specimens was small. In present research, some candidates for the insert material were investigated and also the uniformity of strength on a W/CFC block of the expected size were evaluated. [ABSTRACT FROM AUTHOR]

A DEMOnstration fusion reactor (EU-DEMO) is currently being developed by the EUROfusion consortium, in line with the Horizon Europe research framework programme. This fusion device will be the first on a large scale to generate net electricity. The EU-DEMO reactor is anticipated to undergo a pulsed duty cycle under normal operating conditions, which could shorten the qualified lifetime of the main equipment due to the inevitable induced mechanical and thermal cycling. Furthermore, the plasma control strategy envisaged for the EU-DEMO reactor foresees the potential occurrence of planned and unplanned plasma over-power transients, which might harm the plasma-facing components structure. In light of this, it is essential to dispose of reliable means to predict the thermal–hydraulic performance of the Primary Heat Transport Systems (PHTSs). Given this background, the University of Palermo, in partnership with the DEMO Central Team (DCT), has embarked on a research programme to evaluate the thermal–hydraulic response of the DIVertor Plasma-Facing Components (DIV PFC) PHTS under normal and upset conditions. To this purpose, in order to capture all the relevant geometric, hydraulic and thermal features associated with both ex-vessel and in-vessel components, a detailed finite-volume model has been developed. Next, the analysis of the thermal–hydraulic performance of the DIV PFC PHTS has been conducted both under hypothetical steady-state conditions and during the standard DEMO power cycle. The scope of the activity has been to verify whether the current design of the PHTS is capable of withstanding the pulsating loads it is expected to undergo under nominal conditions. The study has followed a theoretical–computational methodology founded on the use of the thermal–hydraulic system code TRACE. The hypotheses, models and results of the study are presented and commented. • Study on the behaviour of the EU-DEMO Divertor Plasma Facing Components cooling system. • Development of a finite-volume model, based on the requirements of the TRACE system code. • Comparison of model and analytical results and parametric analysis of the ramps. • Validation of the set-up TRACE model to catch the main thermal–hydraulic phenomena. [ABSTRACT FROM AUTHOR]

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. • Information from a limited diagnostic set can be enhanced with previous data. • SOLPS-ITER simulations can augment diagnostic data for emission front control. • The divertor leg emission profile can be reconstructed from a few lines-of-sight. [ABSTRACT FROM AUTHOR]

The EU-DEMO divertor target cooling circuit is equipped with Swirl Tape (ST) inserts to improve its thermo-hydraulic performance in terms of heat transfer coefficient and critical heat flux. Due to the presence of the STs, accurate 3D CFD-based thermofluid-dynamic assessments of the divertor targets cooling circuit require a high computational cost and a laborious pre-processing modelling effort. To this end, a cost-efficient CFD simulation technique based on an equivalent porous medium concept, namely the Virtual Swirl Tape (VST) approach, has been developed. In this work, the mathematical formulation of different VSTs models is presented, and the porous media calibration procedure and validation are shown. This technique enables the reduction of computational costs by decreasing the number of volumes required for a single Plasma-Facing Unit (PFU) assembly cooling channel by a factor of 10, while lowering the calculation time by ≈ 86%. The results obtained show that it is possible to correctly reproduce the friction factor profile and pressure drop of a PFU assembly cooling channel, this latter with errors within 10% considering a wide range of coolant inlet velocities. Some limitations have been observed concerning the VST thermal performance, which is still unsatisfactory and requires further development. The VST approach has been studied using the commercial CFD code ANSYS CFX, coupled with a multi-objective optimization algorithm available in the ANSYS Direct Optimization tool. [ABSTRACT FROM AUTHOR]

• Advanced materials are selected. W-ZrC material exhibits high strength, high ductility, high thermal conductivity, low DBTT, excellent resistance to high thermal-loads and to high flux plasma etching as well as enhanced high-temperature stability. • The mock-up successfully endured the stepped loading (5 - 10 Mw m −2) and cyclical loading (10 MW m −2 1000 cycles). • A thermomechanical measurements for the W-ZrC mono-block mock-up, thermal strain data had been analyzed in detail to explore the mechanism of deformation during thermal cycles. Mono-blocks are important components of plasma-facing components (PFCs) in nuclear fusion engineering. W-ZrC shows superior performance as a novel plasma-facing material (PFM) and may be a better choice as a promising armor material. A W-ZrC mono-block mock-up developed for the divertors of future fusion reactors was tested and analyzed. The displacement, strain, and temperature of the model were obtained using a comprehensive experimental platform combining an electron beam scanning high-heat-flux (HHF) test system with digital image correlation (DIC). The mock-up was successfully subjected to stepped loading (5–10 MW m −2) and cyclical loading (10 MW m −2 for 1000 cycles). The bending effect that caused the mock-up to arch upward was observed. The mock-up was maintained in an elastic deformation state during the entire cyclic process. Linear fitting of the strain data revealed that the strain grew slowly at a growth rate of 0.3 and linear extrapolation found that one of the conditions limiting the lifetime of the mock-up was 24,135 cycles. [ABSTRACT FROM AUTHOR]

• First experimental study on HCFP using helium at prototypical pressure. • Thermal-fluids correlations derived from experimental data. • Thermal-fluids performance estimated at prototypical conditions. Various helium (He)-cooled solid-tungsten (W) divertor concepts have been proposed for long-pulse magnetic fusion energy reactors. Among these concepts, the He-cooled flat plate divertor (HCFP) modules have the largest plasma-facing surface area of ∼0.2 m2. Simulations have shown that the most recent version of the design can withstand heat fluxes as great as 8 MW/m2. Earlier experimental studies of a single shortened HCFP cooling unit with a slot length of 7.6 cm used air at ambient temperature and pressures below 0.6 MPa. Here, we present initial experimental studies of a copper alloy and steel test section modeling a shortened HCFP cooling unit using He at prototypical pressure of 10 MPa, inlet temperatures T i ≤ 200 °C and steady-state incident heat fluxes q ″ ≤ 1.2 MW/m2. Results for Nusselt number Nu as a function of Reynolds number Re were obtained for Re = 1.2×104−3.4×104, and used to develop a Nu (Re) correlation and validate numerical models of the test section using commercial computational fluid dynamics (CFD) software. Simulations with this model are performed to evaluate the effect of the shortened slot. These analyses are used to estimate the thermal-fluids performance of the HCFP under prototypical conditions. [ABSTRACT FROM AUTHOR]

Handling of a large thermal power exhausted from the confined plasma is one of the most important issues for ITER and DEMO. A conventional divertor, which has the closed geometry similar to that of ITER and longer leg of 1.6 m, was proposed for the Japanese (JA) DEMO reactor (Rp/ap = 8.5/2.42 m). A radiative cooling scenario of Ar impurity seeding and the divertor performance have been demonstrated by SONIC simulation, in order to evaluate the power exhaust in JA-DEMO 2014 (primary design with Psep ~ 283 MW) and JA-DEMO with higher plasma elongation (a revised design with Psep ~ 235 MW). The divertor operation with the peak qtarget ≤ 10 MWm−2 was determined in the low nesep of 2–3 × 1019 m−3 under the severe conditions of reducing radiation loss fraction, i.e. f*raddiv = (Pradsol + Praddiv)/Psep, and diffusion coefficients (χ and D). The divertor geometry and reference key parameters (f*raddiv ~ 0.8, χ = 1 m2/s and D = 0.3 m2/s) were so far consistent with the power exhaust concepts in the nesep range, and the revised JA-DEMO design has advantages of wider nesep range and enough margin for the divertor operation. For either severe assumption of f*raddiv ~ 0.7 or χ and D to the half value, higher nesep operation was required for the primary design in order to control the peak qtarget ≤ 10 MWm−2, i.e. the operation window was reduced. Applying the two severe assumptions, the divertor operation was difficult in the low nesep range for the both designs.

The lithium vapor-box divertor is a possible fusion power exhaust solution. It uses condensation pumping to create a gradient of vapor density in a divertor slot; this should allow a stable detachment front without active feedback. As initial explorations of the concept, two test stands which take the form of three connected cylindrical stainless steel boxes are being developed: one without plasma at PPPL, to test models of lithium evaporation and flow; and one for the linear plasma device Magnum-PSI (at DIFFER in Eindhoven, The Netherlands) to test the ability of a lithium vapor cloud to induce volumetric detachment and redistribute the plasma power.The first experiment uses boxes with diameters of 6cm, joined by apertures with diameters of 2.2cm. Up to 1g of Li is placed in one box, which is heated to up to 600°C. The Li evaporates, then flows to and condenses in the two other, cooler boxes over several minutes. The quantity of Li transported is assessed by weighing the boxes before and after the heating cycle, and is compared to the quantity predicted to flow for the box at its measured temperature using a Direct Simulation Monte Carlo code, SPARTA. With good experimental conditions, the two values agree to within 15%.The experiment on Magnum-PSI is in the conceptual design stage. The design is assessed by simulations using the code B2.5-Eunomia. They show that when the H+plasma beam, with ne=4×1020m−3, Te=1.5eV, and r=1cm, is passed through a 16cm long, 12Pa, 625°C Li vapor cloud, the plasma heat flux and pressure on the target are significantly reduced compared to the case without Li. With the Li present, the plasma is cooled by excitation of Li neutrals followed by radiation until it volumetrically recombines, lowering the heat flux from 3.7MWm−2 to 0.13MWm−2, and the pressure is reduced by 93%, largely by collisions of H+with Li0.

A series of dedicated lower hybrid current drive (LHCD) experiments on EAST shows that lithium wall conditioning extends LH current drive and heating up to the line-averaged density of n-e≈4x1019 m−3 for both 2.45 and 4.6 GHz. Current drive at such a high density is crucial for the development of long-pulse non-inductive scenarios on EAST. With lithiation, the LH power injection of 1.5 MW at 2.45 GHz resulted in a drop of loop voltage of ~ 0.3 V, which is a comparable loop voltage drop observed with 1.1 MW at 4.6 GHz. The observed decrease in loop voltage is attributed mostly to the RF heating effect. Another LHCD experiment suggests that lithium wall coating has a more significant impact on the scrape-off-layer (SOL) properties than changes in the Greenwald fraction. LHCD at 2.45 GHz still suffers from a loss of efficiency. Enhanced power ionization in front of the launcher may cause the onset of density-dependent wave instabilities. The rise in the midplane SOL density may also accelerate a transition in the divertor regime, leading to additional ionization and collisional losses in the X-point divertor plasma. Ray-tracing modeling supports that a lower wave frequency is more prone to collisional power loss. The experiments confirm that lithiation is a useful tool to control the SOL plasma, and suggest that density control in front of the launcher may be critical to mitigating power loss mechanisms in the plasma boundary.

Zusammenfassung Der Garchinger Tokamak ASDEX Upgrade ging vor dreißig Jahren in Betrieb. Seither wurde die Anlage, insbesondere ihre Gefäßeinbauten, Plasma‐Heizsysteme und Diagnostiken, beständig erweitert und verbessert. Mit dem Wechsel von einer mit Graphit zu einer mit Wolfram bedeckten Gefäßwand mussten viele Plasmaszenarien überarbeitet werden. Hierbei wurde auch die Kühlung des Divertorplasmas durch zusätzlich eingebrachte Strahlungsverluste entwickelt, die auf ein künftiges Tokamak‐Fusionskraftwerk übertragen werden kann. In vielen Belangen kann ASDEX Upgrade als „Blaupause" eines solchen Tokamak‐Fusionskraftwerks angesehen werden. Eine große Anzahl von Musterentladungen erlauben zusammen mit neu entwickelten Computercodes verlässliche Vorhersagen für ein Kraftwerk. [ABSTRACT FROM AUTHOR]

The results from experimental investigations of chamber internal parts and plasma-facing materials proposed for use in a thermonuclear reactor are reviewed. The thermonuclear reactor internals will experience various heat loads connected with temperature gradients and local exposure dose with the nuclear reaction neutron energy spectrum. The lifetime of the systems will depend on the resistance of materials to neutron and heat loads. Assurance of heat removal is one of serious problems to be solved in implementing a thermonuclear power plant. The components can be cooled either by gas or liquid, including salt solutions and liquid metals. The parameters of coolants, heat-transfer systems, and cooling systems will take values that are still not attainable as of yet. The variety of available design elaborations suggested in the Russian and foreign projects must be substantiated in terms of heat transfer. The parameters of heat and neutron loads will govern the choice of materials for the thermonuclear reactor systems. The international thermonuclear experimental reactor (ITER), which is under construction, will be built using structures, materials, and technologies that have passed the main tests. The ITER cooling systems will operate with forced single-phase convection of turbulent water flow; the heat transfer will be intensified by using inner finning and flow swirling. The thermal protection lining of the plasma-facing chamber's internal parts in the zones of the highest plasma and heat load will be made of tungsten. In elaborating the design of the next-generation demonstration reactor (DEMO), it will be necessary to additionally develop the blanket construction technologies and solve the problem of converting heat into electricity. The materials for the DEMO reactor must be selected with due regard to the high dose of their irradiation by neutrons with the thermonuclear reaction energy spectrum and critically high heat loads experienced by the plasma-facing chamber's internals. It will be necessary to develop and test new materials for constructing the DEMO thermonuclear reactor and solve matters concerned with their commercial-scale manufacture. [ABSTRACT FROM AUTHOR]

Power handling experiments with a special liquid metal divertor module based on the capillary porous system technology were performed in the tokamak COMPASS. The performance of two metals (Li and LiSn alloy) were tested for the first time in a divertor under ELMy H-mode conditions. No damage of the capillary mesh and a good exhaust capability were observed for both metals in two separate experiments with up to 12 MW/m2 of deposited perpendicular, inter-ELM steady-state heat flux and with ELMs of relative energy ~3% and a local peak energy fluence at the module ~15 kJ.m−2. No droplets were directly ejected from the mesh top surface and for the LiSn experiment, no contamination of the core and SOL plasmas by Sn was observed. The elemental depth profile analysis of 14 stainless-steel samples located around the vacuum vessel for each experiment provides information about the migration of evaporated/redeposited liquid elements.

Thermal responses and fatigue behaviors of the tungsten armor of monoblock divertor under steady-state and transient high heat flux (HHF) loads as well as the coupling of the two were simulated and analyzed by the finite element method. It was found that, under the action of thermal cycle loads, the earliest fatigue failure area of tungsten armor is on the surface, and with the increasing of HHF load, the damage area of the armor extends also, while the transient load only causes damage on surface of the armor. The fatigue lifetime of tungsten armor decreases rapidly with the increase of load, whether under steady-state load or transient load. Different transient load under the same steady-state load is researched. When two loads are coupled, the fatigue life is shorter as the load value is high, while the fatigue lifetime is basically unchanged when the transient load is the same under different steady-state loads, that is to say, the steady-state load has little influence on the results, and the transient load during ELMs is the leading factor for the fatigue lifetime.

One of the issues faced by future fusion devices will be high divertor target heat loads. Alternative divertors can promote detachment, flux expansion and dissipation mechanisms to mitigate these heat loads. They have been investigated in several devices including TCV and DIII-D, and will be investigated on MAST-U. To evaluate their effectiveness, accurate target heat flux and power balance measurements are required in these machines. Infrared (IR) thermography is a widely used technique to determine the target heat flux, but is susceptible to surface effects and emissivity in carbon-walled machines. In this work, the effect of plasma exposure on graphite is assessed to understand what may happen in MAST-U. A sample of fine grain graphite, as used on MAST-U, is exposed to 30min plasma exposures, with density ne=6 × 1018m−3and temperature Te=0.08eVas measured by Thomson scattering. During these pulses, the temperature is measured by a medium wave IR camera and is seen to decrease by ≈70 °C over the course of 3h of plasma exposure. Pyrometer measurements suggest that the IR camera data is affected by a change in the surface emissivity. Profilometry confirms erosion of graphite at the tile centre to a depth of ≈100µm, and a larger region of deposition further out, amounting to ≈40µmof material.

Currents structure in the scrape-off layer (SOL) of a tokamak is analyzed. It is demonstrated that poloidal currents measured in the experiments are a combination of several current types of different physical nature. Besides known Pfirsch-Schlüter (PS) currents and thermoelectric currents, so-called PCC (plate closing currents) flowing to/from the divertor plates are also analyzed. The latter close radial currents in the SOL and below/above the X-point. In particular, current flowing to the outer plate in the private flux region (PFR), opposite to thermolelectric current is predicted for the standard single-null configuration and favorable direction of ∇B drift (∇B drift of ions is directed towards active X-point). In addition, a pair of currents to and away from the outer plate should flow. In the single-null configuration they are often masked by a larger thermoelectric current, however for the connected double null (CDN) case, where thermoelectric current is strongly reduced due to less temperature asymmetry, these currents dominate. The suggested physical model is supported by results of simulations performed with SOLPS-ITER transport code. Simulations were done for ASDEX-Upgrade (AUG), L-mode , single-null configurations, and for Globus-M H-mode, both disconnected and connected double null configurations. Results of the simulations are compared with probe measurements for AUG and Globus-M tokamaks, and reasonable agreement has been found.

Reliability and security of the fusion device are severely affected by the failure of the divertor. It is necessary to have a reliable life prediction for heat sink in W/CuCrZr divertor target, especially during long-term high heat flux operation in future fusion devices in which the high temperature creep accelerates its fatigue failure. In this study, a new approach to creep life was applied to creep-fatigue damage accumulation model for CuCrZr heat sink. It is proved that the results of new method for CuCrZr creep life calculation is more consistent with the experimental results than the previous empirical formula. Then the temperature and stress/strain results were calculated for two types of ITER-like monoblock using thermal–mechanical analysis which shows that high temperature creep could not occur on EAST divertor heat sink of current operating conditions. Based on creep-fatigue damage accumulation model, the predicted life of monoblock for DEMO divertor are given under different operating mode respectively. The model provides theoretical basis and guidance for the divertor engineering design of future fusion reactor like CFETR and DEMO in which the high temperature creep in heat sink must be taken into account due to long-term operation.

Recent TCV experiments confirm the predicted formation of an electric potential well, below the magnetic X-point, in configurations with unfavorable Btfield direction (ion ∇Bdrift away from the divertor), that substantially reshapes the typical divertor E×Bflow pattern. The local charge balance ∇⋅jin the private flux region (PFR) of diverted tokamak plasmas has been previously argued to be dominated by parallel and diamagnetic currents. This hypothesis is tested herein in TCV discharges by comparison with SOLPS-ITER simulations, fully accounting for drifts and currents. Simulated parallel currents correctly capture measured current profile characteristics for both targets and both Bt-directions, whereas those omitting drifts fail. It is shown that the resulting parallel currents dictate the electric fields in the PFR for low temperature (detached divertor) conditions resulting in locally negative electric plasma potential in configurations with unfavorable H-mode access.

A new, novel approach to liquid metal plasma facing components called “divertorlets” is presented and accompanied by experiments, simulations, and analysis. The development of a robust and reliable plasma facing component at the divertor is ongoing, with liquid metal divertor concepts gaining interest by showing promise for being able to handle higher heat fluxes as well as improve plasma performance through a reduction in particle recycling. The presented design in this work seeks to address challenges associated with evaporation, operation power, and liquid metal inventory. Divertorlets utilize many adjacent narrow channels with alternating vertical velocity that maintain large flow rates with small velocities at the surface by minimizing the flow path length. Preliminary results using a test stand on LMX-U at PPPL and simulations in COMSOL demonstrate the successful operation and the potential for divertorlets to remove large heat fluxes, with projections made to reactor scale showing the expected system performance.

The transport as well as accumulation of tungsten (W) impurity is one of the critical issues for the W divertor operation. The EAST tokamak with W and graphite divertor targets, provides a good platform for the investigation of W impurity behaviors. In this work, the production, transport and accumulation of W impurities are simulated via the SOLPS-DIVIMP modeling. The plasma background is obtained by the SOLPS simulations, while the W sputtering yield is calculated via empirical formula, and the following W transport is traced using the DIVIMP code. The W impurity source is mainly contributed by the incidence of carbon (C) impurity in this simulation. The distribution of W ions with different charge states is obtained and the transport of the W ions is analyzed by the net forces. The dependences of W impurity on the ne,sepOMP and PSOL are illustrated, which show the divertor plasma has great impact on the W impurity distribution. The increase of ne,sepOMP can suppress the W impurity in the core region with the fixed PSOL, while higher input power can enhance the W concentration significantly. The W impurity can lead to sufficient power radiation due to strong impurity source and high radiation rate in the core region during low density discharge, while the influence becomes weaker as ne,sepOMP raises. The effects of self-sputtering on the W impurity accumulation are also discussed. With the increment of PSOL, the sputtering and self-sputtering of W increase remarkably, leading to the W ions concentration exceed 10−5 in the core–edge interface (CEI). Moreover, the strong correlation between Te at the outer target and W ion density at CEI is observed, showing that Te

Two small liquid metal targets based on the capillary porous structure were exposed to the divertor plasma of the tokamak COMPASS. The first target was wetted by pure lithium and the second one by a lithium-tin alloy, both releasing mainly lithium atoms (sputtering and evaporation) when exposed to plasma. Due to poorly conductive target material and steep surface inclination (implying the surface-perpendicular plasma heat flux 12–17 MW/m2) for 0.1–0.2 s, the LiSn target has reached 900 °C under ELMy H-mode. A model of heat conduction is developed and serves to evaluate the lithium sputtering and evaporation and, thus, the surface cooling by the released lithium and consequent radiative shielding. In these conditions, cooling of the surface by the latent heat of vapor did not exceed 1 MW/m2. About 1019 lithium atoms were evaporated (comparable to the COMPASS 1 m3 plasma deuterium content), local Li pressure exceeded the deuterium plasma pressure. Since the radiating Li vapor cloud spreads over a sphere much larger than the hot spot, its cooling effect is negligible (0.2 MW/m2). We also predict zero lithium prompt redeposition, consistent with our observation.

As a part of Comprehensive Research Facility for Fusion Technology (CRAFT) project, a large superconducting linear plasma testing facility will be built to evaluate candidate plasma-facing materials and component mockups for China Fusion Engineering Test Reactor (CFETR). Powerful plasma source is a key component to produce reactor-relevant particle flux for material irradiation purpose. To develop source technology, a new testing platform is built at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). This platform utilizes conventional water-cooled magnets to sustain a steady-state field strength of > 7000 G and is equipped with a pump train of 4500 m3h− 1, making it capable to evaluate various plasma sources with a fueling rate up to several standard liter per minute (SLM). The ion flux can be measured with a water-cooled target probe. At present, intense plasma beam with an ion flux of > 1023 m− 2 s− 1 has been achieved with a cascaded arc source in this platform. Further testing on the lifetime of source components is ongoing. New cascaded arc type sources based on non-cylinder plasma column concept has been proposed to increase irradiation area and will be tested as well. After finishing source R&D mission, this platform will be upgraded into a plasma-material interaction research facility and open to all researchers in the community. [ABSTRACT FROM AUTHOR]

While high-Z solid plasma-facing components (PFCs) are the leading candidates for reactors, it is unclear that they can survive the intense plasma material interaction (PMI). Liquid metals (LM) PFCs offer potential solutions since they are not susceptible to the same type of damage, and can be "self-healing". Following the Fusion Energy System Study on Liquid Metal Plasma Facing Components study that recently was completed by Kessel et al. (Fusion Sci Technnol 75:886, 2019) a domestic LM PFC design program has been initiated to develop reactor-relevant LM PFC concepts. This program seeks to evaluate LM PFC concepts for a Fusion Nuclear Science Facility (FNSF) or a Compact Pilot Plant via engineering design calculations, modeling of PMI and PFC components and laboratory experiments. The latter involves experiments in dedicated test stands and confinement devices and seeks to identify and answer open questions in LM PFC design. The new national LM PFC program is first investigating lithium as the plasma facing material for a flowing divertor PFC concept. Several flow speeds will be evaluated, ranging from ~ cm/s to m/s. The surface temperature will initially be held below the strongly evaporative limit in the first design; higher temperatures with strong evaporation will be considered in future concepts. Other topics of interest include: understanding of the hydrogen and helium interaction with the liquid lithium; single effect experiments on wetting, compatibility and embrittlement; and prototypical experiments for control and characterization of flowing LM. A path to plasma and future tokamak exposure of these concepts will be developed. [ABSTRACT FROM AUTHOR]

The extreme heat flux anticipated in fusion reactor divertor plasma facing components (PFCs) is perhaps the most challenging technology issue for fusion energy development. Most divertor PFCs are designed based on the maximum steady-state operational limits. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising Li results in NSTX and related modeling calculations motivated the passive and active radiative liquid lithium divertor concepts (RLLD and ARLLD) (Ono et al. in Nucl Fusion 53:113030, 2013; Fusion Eng Des 89:2838, 2014). This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating divertor heat removal with relatively small amount of lithium ~ few moles/s to handle the expected steady-state high divertor heat load. However, in addition to the high steady-state heat flux, the fusion reactor divertor PFCs could also experience significant transient heat flux such as ELMs and/or other magnetic reconnection events which can deposit large transient heat flux onto the divertor PFCs. If unprotected, it could damage the divertor PFC surfaces which could lead to a highly undesirable unplanned shutdown for PFC repair and/or replacement. In this paper, we explore feasibility of LLD and ARLLD concepts in handling such transient heat flux to protect the divertor PFCs from the extreme transient heat flux while maintaining the normal plasma operations. We also suggest a possible implementation technique using inductive pellet injector for the reactor PFC protection from transient heat flux which can be tested on NSTX-U. [ABSTRACT FROM AUTHOR]

In a demonstrational fusion power plant (DEMO), divertor is supposed to protect vacuum vessel and superconducting magnets against neutron flux in the bottom region of the vessel. The vessel is subject to a strict design limit in irradiation damage dose and the magnets in nuclear heating power, respectively. Thus, the DEMO divertor must have the capability to protect sufficiently the vessel and the magnets against neutron flux being substantially stronger than in ITER.In this paper, a first systematic neutronics study for the European DEMO divertor is reported. Results of the extensive assessment of key nuclear loading features (nuclear heating, irradiation damage & helium production) are presented for two optional concepts, namely, dome and shielding liner including minor geometrical variants. The shielding performance of the two competing design options is discussed together with the case of a bare cassette (no shielding), particularly in terms of damage dose compared with the design limits specified for the European DEMO.It was found that both the dome and shielding liner were able to significantly reduce the nuclear loads in the cassette body and the vessel. The maximum damage dose at the end of the lifetime remained subcritical for the cassette body for both cases whereas it exceeded the limit for the vessel under the dome, but only locally on the surface underneath the pumping duct. But, the damage could be reduced below the limit for the vessel by increasing the size of the dome or by deploying the shielding liner. The most critical feature was the excessive damage occurring in the own body of the shielding components where the maximum damage dose in the steel heat sink of the dome and the shielding liner far exceeded the design limit at the end of the lifetime.

We assess the toroidal magnetic field B _t asymmetry in DIII-D due to a misalignment of the toroidal field coils with respect to the poloidal magnetic field coils and vacuum vessel. The peak-to-peak variation of the divertor strike point (SP) radius is measured to be 1 cm, with an n = 1 toroidal pattern. We use the centre of a narrow carbon deposition band on tungsten-coated divertor tiles just inside the outer strike point (OSP) as a proxy for the divertor SP location. The band occurred in a series of reverse B _t discharges with the OSP positioned on the divertor inserts due to strong E × B drift transport of C from the inner to the outer SP through the private flux region. The variation in band radius (and hence the magnetic SP) is a (4.89 ± 0.31) mm shift toward (310 ± 4)° toroidal direction. These measurements agree well with previous measurements of the 3D magnetic field distribution (Luxon 2003 Nucl. Fusion 43 1813), simulations performed by the mafot field line integration code, and recent Langmuir probe measurements in the small-angle-slot (SAS) divertor (Watkins et al 2019 Nucl. Mater. Energy 18 46). Comparison of these measurements in the SAS divertor also indicates that there is the possibility of a tilt (in conjunction with the shift) of the B _t coil field of (0.04 ± 0.07)° towards the toroidal angle of (215 ± 25)°. Previous measurements suggested a field misalignment of (4.6 ± 0.3) mm in the 270° toroidal direction, and a tilt of (0.06 ± 0.02)° toward the 114° toroidal direction, which is similar to the results reported here. These studies will be important for better understanding the radial variation of the toroidal strike line in DIII-D, for designing the new generation of SAS divertor, and for developing an understanding of the impact of error fields on tokamaks with tightly baffled slot divertors.

To discuss a suitable porous structure for helium gas cooling under high heat flux conditions of a nuclear fusion divertor, we first evaluate effective thermal conductivity of sintered copper-particles in a simple cubic lattice by direct numerical heat-conduction simulation. The simulation reveals that the effective thermal conductivity of the sintered copper-particle highly depends on the contacting state of each particle, which leads to the difficulty for the thermal design. To cope with this difficulty, we newly propose utilization of a unidirectional porous tube formed by explosive compression technology. Quantitative prediction of its cooling potential using the heat transfer correlation equation demonstrates that the heat transfer coefficient of the helium gas cooling at the pressure of 10 MPa exceeds 30,000 W/m2/K at the inlet flow velocity of 25 m/s, which verifies that the unidirectional porous copper tubes can be a candidate for the gas-cooled divertor concept.

In the frame of the pre-conceptual design activities of the DEMO work package DIV-1 “Divertor Cassette Design and Integration” of the EUROfusion program, a mock-up of the divertor outer vertical target (OVT) was built, mainly in order to: (i) demonstrate the technical feasibility of manufacturing procedures; (ii) verify the hydraulic design and its capability to ensure a uniform and proper cooling for the plasma facing units (PFUs) with an acceptable pressure drop; and (iii) experimentally validate the computational fluid-dynamic (CFD) model developed by the University of Palermo. In this context, a research campaign was jointly carried out by the University of Palermo and ENEA to experimentally and theoretically assess the hydraulic performances of the OVT mock-up, paying particular attention to the coolant distribution among the PFUs and the total pressure drop across the inlet and outlet sections of the mock-up. The paper presents the results of the steady-state hydraulic experimental test campaign performed at ENEA Brasimone Research Center as well as the relevant numerical analyses performed at the Department of Engineering at the University of Palermo. The test facility, the experimental apparatus, the test matrix and the experimental results, as well as the theoretical model, its assumptions, and the analyses outcomes are herewith reported and critically discussed.

At present, the Kazakhstan Tokamak for Material testing (KTM) project is in progress. The optimization of innovative design elements of the divertor is carried out at the KTM, in addition to studying materials applied for fusion technologies. The design of the lithium divertor module based on lithium capillary porous systems has been developed and tested successfully in real tokamak conditions. In this design, the problem of removing high-density heat flows is solved by using liquid metal coolant based on the Na–K eutectic alloy. The requirements for the improved safety and the divertor design compatibility with other in-vessel water-cooled elements of the tokamak and temperature limitation of the receiving lithium surface to the level <600°C at the heat flows of 10–20 MW/m2 stipulated a new design solution of the divertor experimental module and the use of a totally new coolant—water flow dispersed with gas (gas-water spray). In the paper, the design solutions of the new version of the module, the coolant parameters, and the cooling system layout are described and substantiated. The experimental results on determining the heat transfer coefficient of the coolant based on gas-water spray are considered. [ABSTRACT FROM AUTHOR]

Recently, it was shown that strong electron thermionic emission from material walls could result in the formation of an "inverse sheath," which prevents the flow of cold ions to the wall.[1–3] Such regimes look very favourably from the point of view of plasma–material interactions at the edge of magnetic fusion devices, where the problem of the erosion of plasma‐facing components under ion irradiation is one of the key issues for the development of future magnetic fusion reactors. However, it is not clear whether such regimes are compatible with edge plasma parameters and heat removal requirements in fusion reactors. To address the issue of practicality of the "inverse sheath" regime for edge tokamak plasma conditions, we perform a set of numerical simulations with 2D edge plasma transport code UEDGE[4] for a DIII‐D‐like geometry and magnetic configuration. To describe both "standard" and "inverse sheath" conditions within the framework of the UEDGE code (which does not consider the sheath region per se), at the material surfaces, we apply effective boundary conditions that emulate both "standard" and "inverse sheath" regimes. We demonstrate that, for the same input parameters, spatial distributions of edge plasma parameters corresponding to detached divertor and "inverse sheath" regimes are similar, with only a few minor differences. We discuss the compatibility of "inverse sheath" regimes with core plasma parameters. [ABSTRACT FROM AUTHOR]

In order to study the potential impurity seeding operation regime of the future fusion devices, the first application of the integrated divertor code SONIC to the Ar + Ne mixed‐impurity seeding operation of JT‐60SA steady‐state high‐β plasma has been carried out. In the case, Ne is added to Ar‐only seeding, the separatrix electron density has fell into the desired low separatrix electron density of the scenario. This is mainly because the D+ flow velocity towards the inner divertor has been increased by the Ne seeding. The resultant friction force transports Ar impurities towards the inner divertor region, while impurities are stagnated in the top of scrape‐off layer (SOL) in the Ar‐only seeding case. The higher impurity radiation power in the divertor regions and lower one in the SOL region above the X point have been obtained in mixed‐impurity seeding cases, which show similar tendency as the Ar + Ne mixed‐impurity seeding experiment in JT‐60 U. At the core edge, Zeff has been slightly increased and the radiation power has been decreased as the Ne seeding rate increases. The core plasma/impurity transport has been also evaluated by the TOPICS code using the impurity density at the core edge computed by the SONIC as a boundary parameter. The results show lower Zeff and radiation power, and higher electron temperature in the core in the mixed‐impurity seeding cases. Above possible contributors to the better energy confinement indicate that the mixed‐impurity seeding operation might be more effective than Ar‐only seeding operation. [ABSTRACT FROM AUTHOR]

The decay of the electric potential in the sheath region in tokamak divertors occurs on a scale length on the order of the main ion gyroradius (Chodura sheath) due to magnetic fields lines intersecting the divertor plates at grazing incidence. As a consequence, high‐Z impurities like tungsten ionize within the sheath region in attached plasma conditions. The modification of the electron distribution in the sheath region must thus be taken into account to accurately model ionization and emission of impurities within the sheath region. To that end, an analytical expression of the distribution of the vertical ionization path for impurities sputtered from divertor plasma‐facing components is derived. This expression is then used to estimate the fraction of neutral impurities ionizing within the sheath and the average vertical ionization path, and to derive an effective SXB (the number of ionizations per emitted photon) coefficient which includes the effects of the variation of the electron distribution in the sheath region. These results are applied to tungsten impurities sputtered from divertor plates. It is shown that the SXB coefficient for neutral tungsten is significantly reduced in high‐density attached divertor plasma conditions (ne ≳ 5 × 1013 cm−3) because of the ionization of neutral tungsten well within the sheath region. [ABSTRACT FROM AUTHOR]

Simulations for DIII‐D high confinement mode plasmas with the multifluid code UEDGE show a strong role of poloidal E × B drifts on divertor heat transport, challenging the paradigm of conduction‐limited scrape‐off layer (SOL) transport. While simulations with reduced drift magnitude are well aligned with the assumption that electron heat conduction dominates the SOL heat transport, simulations with drifts predict that the poloidal convective E × B heat transport dominates over electron heat conduction in both attached and detached conditions. As poloidal E × B flow propagates across magnetic field lines, poloidal transport with shallow magnetic pitch angles can reach values that are of the same order as would be provided by sonic flows parallel to the field lines. These flows can lead to strong convection‐dominated divertor heat transport, increasing the poloidal volume of radiative power front, consistent with previous measurements at DIII‐D. Due to these convective flows, the Lengyel integral approach, assuming zero convective fraction, is expected to provide a pessimistic estimate for the radiative capability of impurities in the divertor. For the DIII‐D simulations shown here, the Lengyel integral approach underestimates the radiated power by a factor of 6, indicating that, for reliable DIII‐D divertor power exhaust predictions, full two‐dimensional (2D) calculations, including drifts, would be necessary. [ABSTRACT FROM AUTHOR]

The Chinese Fusion Engineering and Testing Reactor (CFETR) has been performed and is in its engineering design stage. Presently, there are two options allowed for the divertors; the helium-cooled stage and water-cooled stage. The water-cooled one has been focused in this article. The configuration of CFETR divertor module was designed. The cassette design scheme and cassette box inner and outer locking systems were introduced, and the locking system was designed to compliant with the remote-handling (RH) requirements. Two divertor RH design options were provided: one is removed through lower port; the other is targets and baffles removed through upper port; supports and cassette were not removed. The other major research was the water cooling channels’ design and its pressure drop calculation. In addition, the fluid, thermal, and mechanical analyses on the outer target were done. Within 10-MW/m2 heat flux, outer target design can ensure that the temperature for all the components works within their temperature limit, and all the components meet the stress design criteria except the copper alloy pipe. The maximum primary stress induced by pressure on the copper alloy pipe appears in the local position between the two monoblocks, which exceeds the design limit. Protection components are need to be added at the gap of the future design optimization to reduce the local stress. After the thermal-hydraulic analysis on water-cooled divertor cassette, the maximum temperature and the maximum stress of cassette meet the requirements. The results of electromagnetic force analysis shows that the deformation on first wall and transition support is less than 0.5 mm, and the maximum stress is also below its tensile strength values. This article summarized the latest achievements in the design and research of CFETR water-cooled diverter in China. All the results will be of great reference for the CFETR divertor optimal design. [ABSTRACT FROM AUTHOR]

The Materials Plasma Exposure eXperiment (MPEX) is a linear plasma device that will address the plasma–material interaction (PMI) science for future fusion reactors and will enable testing of plasma-facing components (PFCs). It is designed as a steady-state device eventually delivering an ion fluence of up to 1031 m−2 to the target. The device will be designed to handle neutron-activated materials. These capabilities should push the technical readiness level of PFCs up to six for some end-of-life aspects. In order to achieve the relevant plasma conditions, as they are expected in future fusion reactor divertors, MPEX will utilize a novel plasma source concept. This plasma source concept uses a high-power helicon (200 kW, 13.56 MHz), an electron cyclotron heating (ECH) system, which will heat electrons via electron Bernstein wave (EBW) heating (about 400 kW, 70 GHz), and an ion cyclotron heating (ICH) system (400 kW, 6–9 MHz). The physics basis for this plasma source concept including the heating physics and transport is based on experiments on Proto-MPEX. An overview of the experimental results and the physics basis will be given. The physics basis directly translates into functional requirements of MPEX and the conceptual design. The status of the conceptual design of MPEX will be shown. [ABSTRACT FROM AUTHOR]

In a demonstration (DEMO) reactor, mitigation of the large heat load on the divertor target to the below material and engineering limits is a key requirement for operation. Systems modeling is used to design entire fusion power plants and, therefore, has to be able to appropriately capture the divertor challenge. Therefore, it is important to validate these models against comprehensive SOL-divertor simulation codes and experiments. A 1-D divertor model in PROCESS was investigated, compared to the results of 2-D SONIC simulation under the detachment condition. The comparison shows how the 1-D divertor model handles the power loss mechanisms from the outboard mid-plane to the outer divertor target for a DEMO-like condition. The results show good agreement on the calculated value of the total power crossing the separatrix (<5% difference) and the total impurity radiation power $P_{\mathrm{ imp}}$ (<10% difference). However, the 1-D profiles show differences in density and temperature at the upstream of the target (<10 m of connection length to target, corresponding to 10 cm in the poloidal length). One reason for this difference is that the 2-D model calculates impurity transport, which produces a variable impurity fraction along the connection length in the divertor, while the 1-D model uses a single averaged value. The SONIC code also considers physical processes not covered in the 1-D model, such as radial transport in the SOL and divertor region. A scan of $P_{\mathrm{ sep}}$ values in PROCESS for a DEMO-sized machine show that above $P_{\mathrm{ sep}} = 200$ MW, there is a stronger impact on cost and machine size for higher $P_{\mathrm{ sep}}$. [ABSTRACT FROM AUTHOR]

Lasting for already six decades, the current high recycling approach to tokamak fusion is still unable to demonstrate the fusion gain factor $Q_{\text {DT}}=1$ , which is the minimal milestone on the path to fusion energy. Clearly, an alternative approach is needed. Recently, such an approach was proposed as the low recycling regime with plasma pumping by continuously flowing liquid lithium (24/7-FLiLi) and physics of a burning plasma was described in application to the JET tokamak. This article complements the plasma physics by consideration of feasibility and specifics of the low recycling divertor (LRD). Its significant difference from existing high recycling divertors is in the absence of a sheath potential on the open field lines, resulting in small angle impingement to the target surfaces by high energy plasma particles. For the first time, this article presents the design guidance for LRD suitable for JET-like tokamak. Being the only new element in the machine, the LRD with other existing JET facilities, e.g., neutral beam power (NBI) $P_{\textrm {NBI}} = 4$ MW, $E_{\textrm {NBI}} = 120$ keV, opens up the opportunity for demonstration of fusion power $P_{\text {DT}} > 20$ MW, $Q_{\text {DT}} > 5$ and of outstanding tritium burn up, exceeding 7%. [ABSTRACT FROM AUTHOR]

The surface eroding thermocouple (SETC) has proved to be a useful fast local temperature and heat flux diagnostic since its application in DIII-D. The SETC system in DIII-D was upgraded, and seven new SETCs were installed in the small angle slot (SAS) divertor to create a localized poloidal array to provide heat flux profiles with sub-10-ms temporal resolution. In addition, a smaller SETC was developed to increase the spatial resolution and simplify installation in the SAS divertor. The SETC-containing tiles were re-engineered with a double-wide format so that the distance between the SETCs and the tile edges is extended to 10 cm on both sides, giving additional tolerance for tile-to-tile misalignments. Mineral insulated thermocouple cable with a metal sheath has also been implemented to further reduce noise and magnetic pickup. Recent measurements obtained from the SETCs indicate that the newly installed SETC array in the SAS divertor is able to provide important heat flux information to various physics experiments in closed divertors. [ABSTRACT FROM AUTHOR]

The dissipative divertor for the power and particle exhaust in a tokamak relies heavily on strong impurity radiation inside the divertor volume. In the present paper, the influence of the cross-field transport inside the divertor volume on the radiation loss by nitrogen and neon impurity seeding is studied in the SOLPS4.3 2D transport code. The study is carried out in a simplified tokamak geometry with long divertor legs, aimed at amplifying the effect of impurity radiation. It is shown that despite the simple analytical models predicting the enhancement of the low-Z impurity radiation loss with increasing cross-field heat conduction inside the divertor volume, the detailed 2D simulations demonstrate such a trend only if the electron density at the separatrix is used as the principal parameter quantifying the density of the edge plasma. However, if the total amount of the hydrogen isotope particles, which is a more suitable parameter for the edge plasma characterization, is considered, then the cross-field transport shows practically no impact on the impurity radiation in the divertor. It is found also that for a fixed ratio of the impurity to the main plasma content in the edge, two stable solutions can exist in a rather wide interval of the densities of the edge plasma. These solutions are attributed to the different distributions of the seeded impurity between the outer and inner divertors. Jumps between these two stable branches may result in bifurcation-like transitions between the attached and detached divertor regimes. [ABSTRACT FROM AUTHOR]

China Fusion Engineering Test Reactor (CFETR) will be a steady-state fusion power plant using a tritium blanket system. The current challenge is to develop a shielding blanket (SB) system which shields the components from nuclear heat impact. In this article, neutronics issues on the nuclear heat deposition induced by neutrons and photons are studied. It is found that the nuclear heat in the SB modules of breeding regions are 0.04–0.12 W/cc, and it is increased to 0.08–0.115 W/cc in the divertor region. This indicates that the SB concepts in different regions lead to the same effect on nuclear heat deposition. Moreover, it indicates that the nuclear heat trends on the two kinds of particles present a shrunken trumpet distribution in an SB interior and finally could be reduced to the minimum. In particular, the different nuclear heat levels are due to the different materials used in the components, while the steel zones absorb more photon heat than neutron heat and water zones absorb more energy from neutrons than steel. In addition, the thermal-hydraulics issues are discussed based on the computational fluid dynamics (CFD) technology. This article clarifies the nuclear heat distribution issues and also proposes an approach toward heat removal exploration, which would be beneficial to the SB design concepts in the blanket engineering phase later. [ABSTRACT FROM AUTHOR]

To demonstrate the use of embedded thermocouples in new National Spherical Tokamak eXperiment Upgrade (NSTX-U) graphite plasma-facing components (PFCs), a convolutional neural network (CNN) has been trained using the ANSYS simulations to predict the scrape-off layer (SOL) heat flux width, $\lambda _{q}$ , given various machine operational parameters and diagnostic data as inputs. The proof-of-concept CNN was trained on the thermocouple data generated by the approximated NSTX-U heat loads applied to real PFC designs in ANSYS. Once trained, the CNN is capable of high precision reconstruction of parameterized heat flux profiles expected in NSTX-U. In addition, to test the system’s ability to cope with noise and systematic error, pseudonoise was injected into the simulated data. CNN can accurately predict the incident heat flux despite this noise and error. [ABSTRACT FROM AUTHOR]

In this study, global particle balance of helium (He) long pulse discharge with graphite divertor in the Large Helical Device (LHD) and He retention in graphite from lab-scale experiments were compared in order to determine the role of the graphite divertor in He retention on the LHD wall. Global He particle balance analysis was conducted in long-pulse discharges in LHD with only turbo molecular pump. The analysis showed that static He retention was ∼2.9 × 1022 He and the ratio of retained He (= retention over fluence) was ∼0.45%. In the lab-scale study, He retention was measured by Thermal Desorption Spectroscopy (TDS) after ion irradiation or plasma exposure. The ratio of retained He were 0.1 ∼ 0.5% in all conditions, which was well consistent with LHD results. Therefore, it may be concluded that graphite divertor tiles have an important role in He absorption in the LHD wall at the initial phase of discharges.