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Abstract ITER divertor outer vertical target (OVT) consists of 22 plasma facing units (PFUs) and 2 stainless support structures (SSSs) including a complex internal structure for coolant water. Pressure drop and flow rate in each PFU are identified as important parameters for acceptable range in terms of the pumping power of cooling water system and performance of heat removal. In this study, computational fluid dynamics (CFD) and a transient reactor analysis code (TRAC-PF1) which is simpler simulation than the CFD have been performed to estimate the pressure drop and flow distribution in 11 PFUs mounted on an OVT SSS (half of the OVT). Results of both prediction methods showed the proper pressure drop and the homogeneous flow distribution within acceptable range of design requirement. To verify these numerical simulations, the results were compared to experimentally measured values in the flow test. This study demonstrated that the TRAC-PF1 as well as the CFD are applicable tools for further optimization of the divertor OVT cooling structure in view of manufacturing. [ABSTRACT FROM AUTHOR]

Highlights • A new system which stirs liquid metal flow in the magnetic field has been proposed. • The new system generates a wavy flow using the magnetohydrodynamics effect. • A proof-of-principal experiment was conducted under the magnetic field of 5.0 T. • The wavy flow was strengthened with a rise in the magnetic field. Abstract This study proposes ″multi-layered trenches″ as a new liquid metal flow stirring structure for the flowing-type liquid metal divertor to prevent the temperature stratification and the evaporation of liquid metal.Then it investigates the characteristics of the liquid metal flow induced by this structure experimentally and numerically. The proof-of-principal experiment using a torus channel simulating the trenches successfully demonstrated the liquid metal wavy flow where the ratio of meandering component to the main stream was 27% under the magnetic field up to 5 T. The experiment and three-dimensional magnetohydrodynamics flow simulation also proved that magnitudes of the wavy flow and the vorticity increased with a rise in the magnetic field. Moreover, two-dimensional heat transfer model simulation showed the heat transfer enhancement by the wavy flow. Maximum temperature was 26.6 K lower than that in the case without the wavy flow. [ABSTRACT FROM AUTHOR]

Highlights • A calorimeter for a simultaneous measurement of incident heat and current has been assembled. • Energy reflection and secondary electron emission should be considered in heat evaluation. • Basic concept of the CuspDEC divertor concept has been proved in the experiment. Abstract A calorimeter system capable of DC biasing, consisting of a copper target plate, thermocouples (TC), and a copper electrical wire, has been newly assembled and applied for an experimental study of the cusp-type direct energy converter (CuspDEC) divertor concept. The calorimeter is designed to be capable of measuring the temperature of the target plate when the plate is electrically floated or biased. A beam of helium ions with an energy of 2–5 keV is used to test the calorimeter. The calorimeter successfully measured the heat flux and the measured value corresponded well to the theoretical expectations. In order to prove the deceleration of injected charges, which is a basic concept of the CuspDEC divertor, a DC bias voltage is applied to the target plate during beam exposure. As expected in the CuspDEC divertor concept, the suppression of the increase in the pate temperature by the bias voltage has been observed. [ABSTRACT FROM AUTHOR]

Highlights • The radiative divertor with neon-impurity seeding on EAST tokamak is studied using SOLPS. • The impact of neon-seeding location on the divertor plasma is identified. • The impurity distribution and its resulting power radiation are analyzed. Abstract Divertor detachment regime is required for the future tokamak device operation. The radiative divertor with neon-impurity seeding on EAST tokamak is studied by using SOLPS code package in the present study. By comparing the puffing location at dome and upstream, the impact of the neon-seeding location on the divertor plasma is identified, under different puffing duration and rate condition. The impurity distribution and its resulting power radiation are analyzed to explain the reason of impurity puffing location impact. The simulation results indicate that puffing at the divertor region has more advantages than puffing at upstream region. [ABSTRACT FROM AUTHOR]

Abstract Micro/nanoscopic observations on the surface of the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with ITER-like Wall (JET ILW) have been carried out by means of several material analysis techniques. Previous results from the inner divertor were reported for a single poloidal section of the tile numbers 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. The formation of the thick stratified mixed-material deposition layer on tiles 1 and 4, and erosion on tile 3 were identified. This study is mostly focused on the outer divertor: tiles 6, 7 and 8. In contrast to the inner tile, remarkable surface modifications have not been observed on the vertical target (tiles 7 and 8) where sputtering erosion and impurity deposition would have been almost balanced. Only a specific part of tile 6 (horizontal target) located near the exhaust channel was covered with a stratified ("geological-like") mixed-material deposition layer which mainly included Be and Ni with the thickness of ∼2 μm. Special feature of this mixed layer was that a certain amount of nitrogen (N) was clearly detected in the layer. Since the concentration of N varied with the depth position, it could be depended on the amount of that gas puffed for plasma edge cooling during the JET experimental campaign. In addition to the outer divertor tiles, a very interesting feature of the local erosion and deposition effects is reported in this paper. [ABSTRACT FROM AUTHOR]

Abstract Full tungsten divertor is employed for ITER, and the tungsten will be the most potential candidate for the surface material of plasma facing components in fusion reactor after DEMO. Under a Broader Approach (BA) activity, a satellite tokamak device JT-60SA is expected to contribute the investigation of many factors and issues related to the fusion plasma and components. JT-60SA has a future plan to transfer the surface of plasma facing components from carbon fiber reinforced carbon composite (CFC) to tungsten, and investigate the interaction of plasma and tungsten. For this plan, joining and coating technologies of tungsten on CFC is an important issue. Present research developed a fabrication method of the W/CFC clad plates and investigated the mechanical, microstructural and thermal properties of them. [ABSTRACT FROM AUTHOR]

Highlights • Provides loads and simulations for disruptions and ELMS in FNSF. • Thermal loads are smaller than for ITER or a commercial device, so effects of the transients are less severe. • Electromagnetic loads due to disruptions are lower than in ITER and no permanent damage is expected. Abstract Previous work has addressed the thermomechanical and electromagnetic effects of ELMs and disruptions on plasma-facing components in conceptual, commercial fusion reactors. Given that the proposed Fusion Nuclear Science Facility (FNSF) will have different geometry and plasma parameters, it is necessary to address similar issues in that context. Hence, this paper presents two key results: the thermomechanical response of the first wall and divertor to expected FNSF ELMs and disruptions, and the electromagnetic response of all structures inside the FNSF vacuum vessel to a plasma current quench. Structures considered in the electromagnetic analyses include the vacuum vessel, blankets, structural ring (which supports many of the internal components), and all internal shells used for plasma stability and control. The first wall is a ferritic steel structure with a thin tungsten coating and the divertor is all tungsten. Properties used for the divertor are those of pure tungsten, though it is quite possible that some alloy or nanostructured material would be employed at some point. Both the first wall and divertor are gas-cooled, so coolant pressures are quite high (8–10 MPa). The finite element code ANSYS is used to carry out all analyses and parametric studies are employed in order to identify parameters for which uncertainties will have a significant effect on the predicted performance. The ELM and disruption loads are found to be substantially smaller than those expected in a commercial reactor, so the component designs have more margin. [ABSTRACT FROM AUTHOR]

Abstract Diffusion boning is a preferred method to join tungsten and steel for the manufacturing of some components in the fusion technology. However, solid state diffusion bonding encounters many limitations in producing the W–steel modules because it is not tolerant of joints of variable width, and moreover, its reliability is highly sensitive to surface cleanliness and roughness. Therefore, the use of liquid phase forming interlayer is required to eliminate the need for high surface cleanliness and low surface roughness to ensure good contact and diffusion between the bonded surfaces, overcome the large differences in the physical properties of W and steel. In this work, an innovative method is applied for bonding W to steel using Ti/Cu/Ti eutectic forming interlayer. W–steel joint circular sleeve is prepared by a hot isostatic pressing diffusion bonding method at 1050 ℃ for 1 h under 100 MPa pressure. The microstructural examination and mechanical properties evaluation of the joints show that the bonding of W to steel is successful. Investigations of the bonded specimens show a hard reaction layer containing TiC at the interface between Cu and steel, which can weaken the bonding interface. [ABSTRACT FROM AUTHOR]

Exhaust the very demanding incident heat flux of about 10 ∼ 20 MW/m 2 concentrated on the narrow area of the divertor target is one of the most challenging requirements in the fusion DEMO reactor. The divertor has to remove the plasma power within allowable temperature ranges of the comprising materials such as armor, heat sink, and coolant. To satisfy the requirement, the optimized design of the divertor is significant as well as the selection of the comprising materials. In the K-DEMO (Korean fusion demonstration reactor), the target peak heat flux of 10 MW/m 2 was set in the steady-state operation and a water-cooled divertor concept employing the tungsten monoblock has been primarily considered. RAFM (reduced activation ferritic martensitic) steel has been considered the primary candidate as heat sink material and CuCrZr has been the second best plan in the K-DEMO divertor. The purpose of this study is to derive optimum design for two design options: RAFM and CuCrZr. To carry out the design optimization based on thermo-hydraulic and statistical analyses, input parameters such as top thickness and lateral thickness of monoblock and thickness of coolant tube and output parameters such as maximum temperatures of monoblock and coolant tube are defined. In the parameter correlation analysis, regression equations defining the relation between the input and output parameters were derived with statistical values. Finally, the optimized designs for CuCrZr and RAFM models satisfying objectives and constraints were derived by the response surfaces optimization. The optimized designs applying the derived design parameters showed the designs satisfy the thermal requirement. [ABSTRACT FROM AUTHOR]

A heat sink module has been developed for high-heat flux applications in the divertor region of a near term DEMO divertor. The concept employs water jet impingement, in a cascading flow path, together with CuCrZr as the structural material. This fluid-flow architecture maximises the heat transfer coefficient obtained for a given coolant flow rate. Combined with high operational pressures, it facilitates increased coolant outlet temperatures. A conjugate CFD study was performed to compare its performance to the ITER divertor under a representative heat flux boundary condition of 10 MW/m 2 . A transient thermochromic liquid crystal technique was used to experimentally validate the CFD heat transfer data, using air as the working fluid at matched Reynolds numbers. These results showed that area averaged heat transfer coefficients could be accurately predicted by the CFD to within ∼12%. [ABSTRACT FROM AUTHOR]

High-confinement mode is a very prominent operation style for future fusion device due to its unique advantages. However, the conjuncted edge localized modes (ELMs) are very difficult to control so that divertor plates are very prone to suffer both stationary high heat flux (HHF) loads of long-pulse operating mode and transient shock loads of ELMs. Most previous researches focus on degradation of plasma facing material (PFM), however, as a layer joining PFM and cooling tube, the soft copper interlayer suffers concentrated thermal stress loads due to mismatched thermal expansion of PFM and cooling tube. Its thermal fatigue behavior under such coupled loads is also of great significance to structural safety of divertor component. With such a motivation, the reduction effects on fatigue life time of a typical interlayer of monoblock divertor under series of coupled HHF and ELMs shock loading conditions are investigated. It is found that: (1) The transient shock feature of ELMs loading is propagated into interlayer with less sharp pattern. The increase of damage induced by coupled ELMs loading is limited in single cycle, while the accumulated damage of multiple consecutive coupled loading cycles is increased nonlinearly. (2) Under the coupled HHF and ELMs loading, the fatigue life time of interlayer is generally decreasing. The magnitude of decrease is increasing nonlinearly with the magnitude of ELMs peak and averaged heat flux. (3) For three characteristic parameters of ELMs shock loading such as frequency, duration and peak heat flux, the peak heat flux and frequency are two parameters more sensitive to determine coupled reduction effects on fatigue lifetime of the interlayer, while for high frequency case, time averaged heat flux takes the lead. [ABSTRACT FROM AUTHOR]

Dissimilar W/steel metals were successfully joined by diffusion bonding process with the help of a Ti/Ni composite interlayer. The effect of bonding temperature and holding time on interfacial microstructure and mechanical properties of the bonded joints were investigated. Metallographic and compositional analyses show that no intermetallic compound was observed at the W/Ti and Ni/steel interfaces, but Ti 2 Ni, TiNi and TiNi 3 were formed at the Ti/Ni interface at 850–1050 °C for 0.5–2 h. Tensile test showed that variations in the strength of the joint were strongly related to the intermetallic compounds in the diffusion zone of the Ti/Ni interface. The maximum tensile strength of ∼267 MPa was obtained for the joint diffusion bonded at 950 °C for 1 h, and the failure took place at both the W/Ti interface and W substrate. With further increase of the joining temperature, the holding time or both, the joint strength dropped as a function of the increase in the width of Ti–Ni intermetallic compounds, and the failure occurred at intermetallics layers of the Ti/Ni interface. [ABSTRACT FROM AUTHOR]

SOLPS modelling has shown that divertor plasma detachment occurs at a lower upstream separatrix density in the more closed DIII‐D upper divertor than the open lower divertor, demonstrating the utility of the divertor closure in widening the range of acceptable densities for adequate heat handling. To achieve reduced heat flux and erosion at the plasma‐facing components, future devices will need to operate in at least partially detached divertor conditions . Two‐dimensional fluid plasma models coupled to Monte Carlo neutral transport simulations, such as SOLPS, have been widely used to predict the onset of detachment. In modelling the DIII‐D discharges, the cross‐field transport coefficients are constrained to reproduce the experimental upstream profiles. The closed divertor has been modelled with the same input parameters of the open divertor, allowing a direct comparison of the target conditions in both geometries. SOLPS simulations indicate that a higher molecular density correlates strongly with lower electron temperatures. The increased closure of the upper divertor improves the trapping of neutrals, thereby reducing the power density deposited at the target and facilitating detachment, in agreement with experimental observations. [ABSTRACT FROM AUTHOR]

We employ the SOLPS5.1 code to analyse different impurity choices and injection methods as possible drivers for highly radiative scenarios in the European DEMO (EU‐DEMO). We aim at assessing the existence of a suitable parameter region to safely operate the divertor in H‐mode discharges. It turns out that such an operational region exists, and that puffing is strongly preferred to pellet as the impurity injection method. It also appears that many different impurity mixtures can meet the divertor survival requirements, with a low level of W sputtering. This provides an additional degree of freedom, which will be exploited in the future to optimize the overall reactor performance. [ABSTRACT FROM AUTHOR]

Predictions for the operation of tokamak divertors are reliant on edge plasma simulations typically utilizing a fluid plasma code in combination with a Monte Carlo code for neutral species. Pilot‐PSI is a linear device operating with a cascaded arc plasma source that produces plasmas comparable to those expected in the ITER divertor (Te ∼ 1 eV, ne ∼ 1021 m−3). In this study, plasma discharges in Pilot‐PSI are modelled using the Soledge2D fluid plasma code coupled to the Eirene neutral Monte Carlo code. The plasma is generated using an external source of plasma density and power. These input parameters are tuned in order to match Thomson scattering (TS) measurements close to the cascaded arc source nozzle. The sensitivity of the simulations to different atomic physics models is explored. It is found that elastic collisions between ions and hydrogen molecules have a strong influence on calculated profiles. Without their inclusion, supersonic flow regimes are obtained with M ∼ 2 close to the target plate. Simulation results are compared with experimental findings using TS close to the target and, in the case of Pilot‐PSI, a Langmuir probe embedded in the target. Comparison between experimental trends observed in a background pressure scan and the simulations show that the inclusion of the elastic collision is mandatory for the trends to be reproduced. [ABSTRACT FROM AUTHOR]

In this paper, we present the initial results of a fundamental study of the impact of the magnetic configuration on the detachment behaviour and heat exhaust in HL‐2M simulation for low power. Starting from typical HL‐2M single‐null shapes, the plasma density at the core–edge interface is varied from 1 × 1019 to 8 × 1019 m−3 to investigate the detachment plasma characteristics. The additional X‐point near the primary X‐point and target is invested in snow flake plus (SF+) and snow flake minus (SF−) configurations. Detachment at the outer target is also studied in these simulations. The degree of detachment and the radiation location are investigated using the SOLEDGE2D simulation results. It is found that advanced divertor configurations result in a deeper detachment under the same plasma density and in a reduction in the detachment threshold. [ABSTRACT FROM AUTHOR]

Simulations of the ARC scrape‐off layer and divertor configuration are carried out using UEDGE to perform the first power‐handling assessment for the divertor design. A convective–diffusive anomalous transport model is implemented in the code to reproduce the radial plasma profiles anticipated for the device at the outer mid‐plane. Levels of exhaust power from the core are varied, and a 0.5% fixed fraction neon impurity seeding is included in investigations to find steady‐state detached solutions. Initial studies employing a super‐X divertor configuration have shown that a stable detached divertor‐operational window exists for reactor exhaust powers in the range of 72–92 MW, which is close to projected levels of exhaust power of 105 MW across the separatrix anticipated for ARC. Under these conditions, the divertor heat flux is fully dissipated by radiation and radial losses to the sidewalls of the divertor channel, leading to acceptable levels of power flux density on plasma‐facing components. Simulations are being extended to study the performance of the X‐point target geometry and to explore the sensitivity to the physics model and input parameters. [ABSTRACT FROM AUTHOR]

In a preliminary design study for a small‐angle slot (SAS) divertor configuration suitable for DIII‐D double‐null advanced tokamak (AT) plasmas, a conceptual “SAS 2” slot divertor configuration has been modelled with SOLPS‐ITER. A balanced double‐null equilibrium, representing an H‐mode AT plasma, is employed, with 5 MW neutral beam input power. In simulations with fully recycling walls and no pumping, an up–down‐symmetric open divertor geometry is compared with the SAS 2 configuration. In density scans without electric or magnetic drifts, the upstream separatrix electron density needed to achieve detachment at both target strike points (ne, sep, det.) is ∼35% lower in SAS 2 than in the open configuration, with ne, sep, det. = 1.66 and 2.59 × 1019 m−3, respectively. With drifts, the same ∼35% difference remains, but detachment density for each configuration is shifted up by ∼15%, giving ne, sep, det. = 1.94 and 2.99 × 1019 m−3, respectively. Cryopumping is modelled for the SAS 2 divertor, and with a pump duct deep in the upper slot on the common‐flux side, detachment density is shifted up by ∼15–20%. Strong pumping in the upper slot delays detachment there compared with the lower slot, regardless of drift direction. This research sets the stage for future modelling that would explore more sophisticated cryopumping configurations, and sensitivity of divertor performance to changes in magnetic balance and input power. [ABSTRACT FROM AUTHOR]

This study investigated the characteristics of arcing initiated on nanostructured fuzzy tungsten (W) surfaces in helium (He) plasmas. Under the same He plasma conditions, the relationship between arc voltage and current was characterized as linear due to plasma resistance across the He plasma from the target surface to the plasma source. Properties of the erosion caused by arcing were investigated by focusing on two parameters: arc current and fuzzy layer thickness. The erosion per charge was significantly dependent on the thickness of the fuzzy layer but did not depend on the arc current. Sample surface analysis revealed that drastic changes in arc spot grouping features were accompanied by the fuzzy layer thickness variation. The ecton model indicated that the erosion as ions does not play a main role, but mechanical destruction by momentum transfer from the explosive electron emission centre to neighbouring nanowires seems to enhance the erosion greatly. [ABSTRACT FROM AUTHOR]

Radiation transport in plasma can be important for a number of problems, for example, first wall‐shielding effects or influence of hydrogen isotope radiation on tokamak edge plasma. Existing calculations in the field employ many simplifying assumptions. For example, the model of radiation transport in one of the most used codes, SOLPS, suggests that photoexcitation can be ignored for all transitions, not including the ground level. We develop a collisional radiative model of radiation transport in a multi‐level system, fully accounting for all transitions between levels. The role of the Stark effect in line shape is taken into account. The validity of some of the approximations made in SOLPS is analysed for detached divertor plasma in DIII‐D‐like tokamak geometry. It is shown that total radiation flux is directed away from the divertor target due to nonuniform plasma temperature and density. [ABSTRACT FROM AUTHOR]

Nowadays, divertors are used in the main tokamaks to control the magnetic field and to improve the plasma confinement. In this paper, we present analytical symplectic maps describing Poincaré maps of the magnetic field lines in confined plasmas with a single-null poloidal divertor. Initially, we present a divertor map and the tokamap for a diverted configuration. We also introduce the Ullmann map for a diverted plasma, whose control parameters are determined from tokamak experiments. Finally, an explicit, area-preserving, and integrable magnetic field line map for a single-null divertor tokamak is obtained using a trajectory integration method to represent toroidal equilibrium magnetic surfaces. In this method, we also give examples of onset of chaotic field lines at the plasma edge due to resonant perturbations. [ABSTRACT FROM AUTHOR]

Monoblock technology with CuCrZr as the structural material is the standard divertor target design for the initial phase of the China Fusion Engineering Test Reactor (CFETR). The design concept with China Low Activation Martensitic (CLAM) steel as the structural material provides another candidate solution for the CFETR divertor target. This paper presents a design study of an optimized monoblock to meet the requirements for CFETR operating conditions, which establishes a good basis for engineering design of CFETR. The monoblock uses tungsten as armour and CLAM steel as the structural material, with a copper interlayer. The operating temperature window of CLAM steel was determined by irradiation embrittlement, softening, and other aging effects. Initial design rules for the W/CLAM monoblock were checked; thermal results by finite element analysis (FEA) show that the design concept is able to withstand a heat flux of 10 MW/m 2 after examining the temperature for each material and the margin to the wall critical heat flux (WCHF) on the tube. Mechanical results indicate that the CLAM steel meets the ITER SDC-IC elastic rules under potential irradiation level of CFETR, irrespective of residual stress. Also the fatigue criteria were checked for unirradiated conditions with a reasonable life time of 16,000 cycles. [ABSTRACT FROM AUTHOR]

The present paper addresses the impact of the divertor option on the nuclear performances of the Demonstration fusion reactor (DEMO). As the effect of the number and size of the divertor has been already evaluated, in this work the focus has been posed on the composition in terms of amount of cooling inside the divertor cassette. Transport responses, as the Tritium breeding ratio (TBR), neutron and gamma fluxes and spectra inside the plasma chamber, as well as activation responses such as shutdown dose rate, decay gamma fluxes and heating have been evaluated for two different blanket concepts of the future European DEMO reactor: DCLL and WCLL. Three different divertor compositions have been tested demonstrating the importance of this component not only locally but in the global radiation field. The transport analysis has been performed with the Monte Carlo code MCNP5 and the JEFF3.1.1 and JEFF3.2 data libraries. The activation responses calculated using Advanced D1S method have been recently assessed and summarized in the present paper. [ABSTRACT FROM AUTHOR]

Heat load calculations have indicated the possible overloading of the ends of the water-cooled divertor facing the pumping gap beyond their technological limit. The intention of the scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper is divided into six modules of four plasma facing components (PFCs); each module has four PFCs hydraulically connected in series by two water boxes (inlet and outlet). A full-scale prototype of one module has been manufactured. Development activities have been carried out to connect the water boxes to the cooling pipes of the PFCs by tungsten inert gas internal orbital welding. This prototype was successfully tested in the GLADIS facility with 17 MW/m2 for 500 cycles. The results of these activities have confirmed the possible technological basis for a fabrication of the water-cooled scraper. [ABSTRACT FROM AUTHOR]

In 2015, the optimized stellarator Wendelstein 7-X stellarator (W7-X) started with operation. The main objective of W7-X is the demonstration of the integrated reactor potential of the optimized stellarator line. An important element of this mission is the achievement of high heating power and high confinement in the steady-state operation. The approach to this mission is the following three steps. First, plasmas were produced in a limiter configuration [operation phase (OP 1.1)], then a test divertor unit is being installed (temporary divertor unit) for the next campaign, OP 1.2, before the full steady-state capability will be achieved implementing active cooling of all in-vessel components and a steady-state high heat-flux divertor. In December 2015, the first helium plasma was generated using electron cyclotron resonance heating (ECRH), in February 2016, the working gas was switched to hydrogen. The first OP (OP 1.1) was successfully finished in March 2016. At the end of OP 1.1, the discharge duration was close to 6 s, the limit for the integrated heating power was increased to 4 MJ and electron temperatures of ~10 keV were achieved. Due to the low densities in the range of 1019 cm−3 and the pure electron heating by ECRH, the ion temperatures reached only 2 keV. At present, W7-X is undergoing the next completion phase, including the installation of the test divertor unit, the installation of the carbon tiles on the inner plasma vessel wall, an upgrade of existing diagnostics, and the installation of new diagnostics. This paper discusses the first operational phase, lessons learned and implemented, and the status before the start of the second OP (OP 1.2). [ABSTRACT FROM AUTHOR]

The steady state with duty cycle time 0.3–0.5 of Chinese fusion engineering testing reactor (CFETR) is a big challenge. To realize this goal, the divertor should be reliable. At the preconceptual phase of CFETR design, reliability engineering is engaged to determine how reliable of divertor is and how much level of reliability of divertor should be. The plasma facing components (PFCs), base plate, and support frame have been considered during the reliability analysis for the divertor modules. The relation of the reliability between a divertor system and its modules has been analyzed. The relations between steady-state operation with duty cycle time 0.3–0.5 and the reliability of divertor is studied in this paper. How the reliability of components, modules, and divertor system affects the duty cycle time of CFETR is also analyzed. To meet the requirement of duty cycle time, the lowest reliability level of divertor system is 0.3679 and its mean time to fail is only 1 year. The divertor system with such low reliability requires that each module should have a relative high reliability of 0.9794, and then, the PFCs should have higher reliability of 0.9930. These results are then must become the requirements for the engineering design of CFETR divertor in the future. [ABSTRACT FROM AUTHOR]

The challenge of integrated exhaust consistent with the other requirements in DEMO and power plant class tokamaks (ITER-like and alternative DEMOs, Fusion Nuclear Science Facility approaches) is well-known and the exhaust solution is likely to be fundamental to the design and operating scenarios chosen. Strategies have been proposed such as high main plasma radiation, but improved solutions are sought and will require revised research methodologies. While no facility can address all the challenges, the new MAST Upgrade tokamak enables exploration of a wide range of divertor plasma aspects in a single device and their relation with the core plasma (e.g., access to H-mode) and in particular the development of fundamental understanding and new ideas. It has a unique combination of closed divertor, capability of a wide range of configurations from conventional to long leg (including Super-X), and fully symmetric double null (plasma and divertor structures). To extrapolate to DEMO and power plant scale devices where full integrated tests in advance are not feasible yet different physics mechanisms may dominate, theory-based models are likely to be essential for confident performance prediction, optimization, and a “qualification” of the concept. Development and validation of such models is at the heart of the program around MAST Upgrade. Amongst the many areas to be explored, there will be a strong focus on the closely coupled topics of plasma detachment and cross-field transport mechanisms (e.g., plasma filaments), key ingredients of effective and reliable protection of the plasma-facing components at DEMO scale. [ABSTRACT FROM AUTHOR]

Experimental advanced superconducting tokamak (EAST) is a D-shaped full superconducting tokamak device. Plasma-facing components (PFCs) of EAST lower divertor are most likely to be damaged during the operation, which become fail easily and have to be replaced and repaired frequently. To evaluate whether the divertor is able to reach the design objective, an assessment of technical risks by means of reliability, availability, maintainability, and inspectability (RAMI) for PFCs of EAST lower divertor has to be performed. A functional analysis is carried out in a bottom–up approach, which is described using the IDEFØ methodology. Based on the functional analysis, a failure mode, effect, and criticality analysis is performed to evaluate potential causes of failures and their effects. Criticality charts highlight the risk levels of different failure modes with regard to the probability of their occurrence and the impact on the operation, and actions are taken to mitigate the major risks. According to reliability block diagrams, the availability and the reliability of functions are calculated under stipulated operating condition. The RAMI analysis results provide the reference for the coming upgrade of EAST lower divertor and will be further optimized as the upgrade of the lower divertor. [ABSTRACT FROM AUTHOR]

Experimental advanced superconducting tokamak (EAST) device is a D-shaped full superconducting tokamak with actively water cooled plasma-facing components. Before this upgrade, three generations divertors, which are steel divertor, carbon divertor, and international thermonuclear experimental reactor-like upper divertor, have been applied, respectively. To achieve long-pulse and high- $\beta $ H-mode plasma, new plasma configurations and shapes have been modeled and optimized in 2016 on EAST. The new geometry of lower divertor heavily relies on numerical simulations of the plasma in EAST. The new divertor is designed to fit the high- $\beta $ H-mode plasma and endure the heat flux up to 10 MW/m2. To achieve this goal, the lower carbon divertor will be replaced in EAST, which is now in conceptual design phase. Compared to the outer target side, the inner target side is prone to plasma detachment. First, the geometry and physics of the divertor should be better designed with the advanced physical operation mode. Second, the divertor should be in advanced geometry and high efficient cooling structure. Third, considering balance the detachment between inner target side and outer target side. This paper mainly introduces the research progress of the fourth-generation tungsten divertor in EAST. [ABSTRACT FROM AUTHOR]

A flush-mounted Langmuir probe system has been built on the lower graphitic divertor targets in the experimental advanced superconducting tokamak (EAST) in 2016, which is revised from the previous divertor Langmuir probes, aimed to reduce the erosion from high-energy particle and strong heat flux on the probe surfaces exposed in the plasma, and explore new structure applications. The theories of single and triple flush-mounted probes relevant to their dimensions and the plasma parameters in the EAST divertor region are reviewed. During the 2016 EAST campaign, the flush-mounted probe system measured the plasma parameters by using single and triple probes, respectively, to obtain plasma parameters. The results were compared with the previous measurements by the domed probe in the similar condition. The results showed a good agreement between measurements using different probe geometries and techniques. This demonstrated the reliability of the flush-mounted probe system as a diagnostic tool in the EAST divertor. Optimization of these Langmuir probe system for the next generation of EAST lower divertor, which will be upgraded to a full tungsten divertor with active water cooling, is also discussed. [ABSTRACT FROM AUTHOR]

The coating of tokamak walls with thin layers of lithium has been demonstrated to reduce plasma recycling from the plasma-facing surfaces and to improve overall plasma performance. These effects, including reduced divertor $D_{\alpha }$ emission, the elimination of edge-localized modes, and increased energy confinement have been observed in multiple experiments when lithium coatings are applied before plasma discharges. However, this coating technology does not extrapolate to future long-pulse devices, since the lithium coatings will be passivated by the continual plasma flux onto the surface. In order to provide active conditioning capability, a new technology has been developed that is capable of injecting lithium powder into the scrape-off layer plasma during plasma discharges, where it quickly liquefies and turns into an aerosol. The use of this “lithium dropper” is under study at the Experimental Advanced Superconducting Tokamak (EAST), where the potential benefits of real-time wall conditioning via lithium injection are being tested. Here, we present an analysis of the recycling characteristics during EAST experiments testing active lithium injection in order to assess recycling reduction and control. Lithium aerosol was injected from the top of the machine, with one system dropping lithium near the $X$ -point and another into the low-field side divertor leg. The injection of lithium into the SOL reduced divertor recycling, as evidenced by reduced $D_{\alpha }$ emission with ion flux measured by probes relatively unchanged. This effect is strongest in the active divertor, confirming the lithium is transported to strongly plasma-wetted areas. Quantitative analysis of the recycling changes using the SOLPS edge plasma and neutral transport code indicated a ~20% reduction in recycling coefficient with lithium injection. [ABSTRACT FROM AUTHOR]

The HL-2M divertor consists of a flat-type carbon fiber composite (CFC)/CuCrZr component and a cassette structure. The CFC/CuCrZr component is made of a water-cooled CuCrZr heat sink armored with carbon CFC tiles. The CFC surface was modified by using Cr slurry to improve its wettability to copper. The Cr layer was deposited by a simple screen-printing technique and then sintered at 1300 °C in a vacuum furnace. An oxygen-free copper (OFC) buffer layer was cast on the modified CFC surface in order to mitigate the internal stresses caused by mismatch in the coefficient of thermal expansion of CFC and CuCrZr. The brazing of the cast OFC/CFC tiles to the CuCrZr heat sink was performed by using a silver-free brazing alloy. Microstructural observations and shear strength tests demonstrate high quality of joining for the CFC/CuCrZr joints. The brazed CFC/CuCrZr mock-up experienced cyclic loading tests of 7 MW/m2 for 1000 cycles with no degradation of heat removal, which meets the requirement of HL-2M divertor. [ABSTRACT FROM AUTHOR]

A number of experimental studies of melt motion and droplet ejection caused by pulsed plasma load include the measurements of the shape of surface after the solidification of target. The measured shape may be different from the one during the heating stage because of melt motion. In present paper the evolution of this perturbations is treated as capillary waves on the melt surface. The dispersion relation for capillary waves taking into account viscosity and limited depth of liquid was used. The numerical estimations for the melt surface behavior are done for tungsten samples irradiated at BETA facility. [ABSTRACT FROM AUTHOR]

The shaping of the ITER divertor vertical targets has been refined as a consequence of manufacturing and engineering considerations during the prototype manufacturing activities. In this paper, the optimized ITER divertor design is presented together with design validation by 3D field line tracing calculation and thermo-mechanical analysis by finite element calculations. Furthermore, the reduction of W monoblock armour thickness to 6 mm is also discussed. [ABSTRACT FROM AUTHOR]

Here we elucidate the mechanisms of plastic deformation and fracture of tungsten laminated composites. Our results suggest that the mechanical response of the laminates is governed by the plastic deformation of the tungsten plies. In most cases, the impact of the interlayer is of secondary importance. Severely cold-rolled ultrafine-grained tungsten foils possess exceptional properties in terms of brittle-to-ductile transition (BDT), toughness, and tensile ductility. The motivation for investigating laminated composites is to determine whether a bulk material can be made that retains the ductility of the thin tungsten foils. In this paper we analyse W-AgCu, W-Cu, W-V, and W-Pd laminates in their as-produced and annealed conditions (e.g. 10, 100 and 1000 h at 1000 °C (1273 K) in vacuum). The analyses comprise (i) the mechanical characterisation by means of three-point bending (damage tolerance), Charpy impact (BDT), and tensile tests (total elongation to fracture) as well as (ii) the in-depth analyses of the microstructure by means of scanning electron microscopy (SEM), transmission electron microscopy (TEM), and Auger electron spectroscopy (AES). W-Cu laminates (60 vol% W) show 15.5% total elongation to fracture in a tensile test at room temperature. Furthermore, the BDT of tungsten laminated composites occurs at a temperature that is several hundreds of Kelvin lower than the BDT temperature of the pure tungsten bulk counterparts. Finally, we present the successful fabrication of a 1000 mm long W-Cu laminated pipe and show its high heat flux performance. Fabrication studies of high heat flux components made of tungsten laminates, in which the laminates are used either as heat spreaders or structural pipes, are presented. [ABSTRACT FROM AUTHOR]

A SiC/SiC composite is a candidate for the structural material of fusion reactors after Demonstration Power Plant (DEMO). Tungsten will be employed as an armor of SiC/SiC constructed divertor, the developments of fabrication method and optimization of thermal property of them are the important issues. Present research produced W and SiC/SiC clad plates using a diffusion bonding method and tried to investigate the thermal characters of the interphase between W and SiC/SiC composites. The thermal conductivities of W/SiC clad, SiC/SiC and tungsten plates were characterized by a laser flash method at temperatures up to 773 K. The thermal data was approximated by functions and applied as material data to an analysis by finite element method (FEM). The simple FEM model of the divertor suggested that the existence of the interfacial phase with low apparent thermal conductivity of the divertor system under the heat flux exposure of 5 MW/m 2 may cause fracture of the system due to thermal stress. The results show the importance of the improvement of the interphase to enhance the thermal properties and the reactor design to reduce thermal flux for the divertor surface. [ABSTRACT FROM AUTHOR]

Digital twin techniques enhance traditional engineering analysis workflows of existing systems when a realistic evaluation of a component under complex operating conditions is required. During preparation, commissioning and operating phases, components can be virtually tested by using validated numerical models, operational expertise, and experimental databases. Three complementary applications have been developed under this approach. The numerical models used for the divertor tiles are based on continuum mechanics formulations. Their loading conditions are defined using the current physics and engineering understanding of a combination of experimental measurements. The aim of these tools is to increase operational range, reliability, and predictability of the JET divertor. [ABSTRACT FROM AUTHOR]

During the qualification program of the tungsten divertor vertical targets, the tested mock-ups (250 tested tungsten monoblocks in total) successfully demonstrate their thermal performances and structural integrity. However, some of the tested monoblocks, in average 30%, showed macro-cracks in the tungsten. This paper presents the results of 3D elastic-plastic thermo-mechanical analysis of the tungsten monoblock under stationary thermal loads for two cases of tungsten material properties: (1) stress-relieved and (2) recrystallized. The comparison pointed out that the recrystallized tungsten monoblock accumulated more damages at the loaded than stress-relieved tungsten. In addition, recrystallization may lead to early development of cracks on the monoblock either due to progressive deformation or fatigue while the non-recrystallized monoblock has a much smaller probability to develop cracks, as long as the exposed surfaces are free from defects. [ABSTRACT FROM AUTHOR]

A new liquid metal divertor system named the REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) is proposed for the helical fusion reactor FFHR-d1. The REVOLVER-D is composed of molten tin shower jets stabilized by internal flow resistances of wire/tape/chain. These shower jets are inserted into the ergodic layer surrounding the main plasma. Tin is selected as the liquid metal because of its low melting point, low vapor pressure, low material cost, and high safety. The liquid metal pumps, cryopumps, and turbo molecular pumps are installed in the central vacuum vessel connected to the main vacuum vessel via 10 inner ports equipped with maze neutron shields. Central solenoid coils made of high-temperature superconductors are installed inside the central vacuum vessel to shield the pumps from the strong magnetic field. The REVOLVER-D has a good possibility to satisfy important characteristics required for the divertor system in a fusion reactor, that is, high heat load tolerance, high maintainability, sufficient vacuum pump speed, high level of safety, and a small amount of radioactive wastes. [ABSTRACT FROM AUTHOR]

Since the signatures of ITER divertor Procurement Arrangements, material purchases, process qualification as well as manufacturing of full-scale prototypes have progressed. This paper provides a brief summary of the ITER divertor materials, the requirements for these materials, and the requirements for manufacturing and inspection of the divertor components. Experiences to be acquired through the prototype manufacturing activities are also discussed. [ABSTRACT FROM AUTHOR]

Considering that solutions for the steady-state power exhaust problem in future fusion reactors (e.g. DEMO) are not provided by present experiments and it is uncertain if they will be provided by ITER, because the expected heat fluxes, as well as the level of neutron irradiation, will be much higher, dedicated work packages are being devoted to this problem within EUROfusion and even a dedicated facility (the Divertor Tokamak Test − DTT) is being proposed in Italy. Among the possible options, a liquid metal (LM) divertor is being considered. The present work aims at developing a simple model of the LM loop in the case of a pool-type divertor, including the most important physical phenomena and allowing to roughly determine the operating range of the system, in terms of surface temperatures and vapor pressures. This work therefore sets a preliminary basis for the conceptual design of a LM divertor for the DTT facility. The model includes the incoming plasma heat load and a basic treatment of the interactions of the Li vapor with the plasma. The reduction of the Li vapor efflux due to ionization by the plasma is also taken into account. The model includes two chambers: a first divertor box, the evaporation chamber (EC), is open towards a second divertor box, the differential chamber (DC), which is in turn connected to the main plasma chamber (MC). The model is used to study the effectiveness of the LM vapor in radiating isotropically the parallel heat flux incoming in the divertor. The results indicate that the presence of the DC allows a significant reduction of the Li vapor efflux towards the MC, which in turn would imply a lower contamination of the core plasma. [ABSTRACT FROM AUTHOR]

During the operation of ITER, protective coatings of the divertor and the first wall will be exposed to significant plasma heat loads which may cause a huge erosion. One of the major failure mechanisms of metallic armor is diminution of their thickness due to the melt layer displacement. New experimental data are required in order to develop and validate physical models of the melt layer movement. The paper presents the experiments where metal targets were irradiated by a plasma stream at the quasi-stationary high-current plasma accelerator QSPA-T. The obtained data allow one to determine the velocity and acceleration of the melt layer at various distances from the plasma stream axis. The force causing the radial movement of the melt layer is shown to create an acceleration whose order of magnitude is 1000g. The pressure gradient is not responsible for creating this large acceleration. To investigate the melt layer movement under a known force, the experiment with a rotating target was carried out. The influence of centrifugal and Coriolis forces led to appearance of curved elongated waves on the surface. The surface profile changed: there is no hill in the central part of the erosion crater in contrast to the stationary target. The experimental data clarify the trends in the melt motion that are required for development of theoretical models. [ABSTRACT FROM AUTHOR]