[en] In this work, Wendelstein 7-X (W7-X) is the world’s largest superconducting nuclear fusion experiment of the optimized stellarator type. In the first Operation Phase (OP1.1) helium and hydrogen plasmas were studied in limiter configuration. The heating energy was limited to 4 MJ and the main purpose of that campaign was the integral commissioning of the machine and diagnostics, which was achieved very successfully. Already from the beginning a comprehensive set of diagnostics was available to study the plasma. On the path towards high-power, high-performance plasmas, W7-X will be stepwise upgraded from an inertially cooled (OP1.2, limited to 80 MJ) to an actively cooled island divertor (OP2, 10 MW steady-state plasma operation). The machine is prepared for OP1.2 with 10 inertially cooled divertor units, and the experimental campaign has started recently.The paper describes a subset of diagnostics which will be available for OP1.2 to study the plasma edge, divertor and scrape-off layer physics including those already available for OP1.1, plus modifications, upgrades and new systems. In conclusion, the focus of this summary will be on technical and engineering aspects, like feasibility and assembly but also on reliability, thermal loads and shielding against magnetic fields.

[en] At DIII-D, lithium granules were radially injected into the plasma at the outer midplane to trigger and pace edge localized modes (ELMs). Granules ranging in size from 300 to 1000 microns were horizontally launched into H-mode discharges with velocities near 100 m/s, and granule to granule injection frequencies less than 500 Hz. While the smaller granules were only successful in triggering ELMs approximately 20% of the time, the larger granules regularly demonstrated ELM triggering efficiencies of greater than 80%. A fast visible camera looking along the axis of injection observed the ablation of the lithium granules. We used the duration of ablation as a benchmark for a neutral gas shielding calculation, and approximated the ablation rate and mass deposition location for the various size granules, using measured edge plasma profiles as inputs. In conclusion, this calculation suggests that the low triggering efficiency of the smaller granules is due to the inability of these granules to traverse the steep edge pressure gradient region and reach the top of the pedestal prior to full ablation

[en] The behavior of 16Cr3Al ODS steel (oxide dispersion strengthened steel), widely employed structural fusion material, under high-intensity laser radiation with intensity up to 10(14)W/cm(2) was investigated in air, helium and vacuum surrounding. Employed system was 65 fs laser at 804 nm, with applied pulse energy up to 5.25 mJ. Morphological effects were studied - cracking, crater parameters (depth, cross-section), LIPSS (laser-induced periodic surface structures) formation at the crater periphery, hydrodynamic effects, as well as chemical variations on the surface. Ablation thresholds were also determined for all three ambiences (for 100 applied pulses), and they were 0.30 J/cm(2), 0.23 J/cm(2) and 0.39 J/cm(2) in air, helium and vacuum, respectively. Plasma occurred in all experiments and it was most prominent in vacuum due to strongest laser-material coupling.

[en] Reduction of hydrogen content in deuterium-fueled fusion plasmas is important not only to avoid diluting reacting core deuterons but also so as to allow the optimization of hydrogen-minority-heating efficiency in those fusion devices which employ ion-cyclotron-radio-frequency heating systems. In EAST, the amount of hydrogen released from plasma-facing components has been shown to depend strongly on both their composition and their temperature. In this paper, as measured by thermal desorption spectroscopy, the hydrogen inventory in graphite – used in EAST as lower divertor material – has been determined to be >25 times larger than that of tungsten which comprises the upper divertor. This difference in hydrogen inventory is attributed mostly to the intrinsically porous nature of bulk graphite. Thus the main source of hydrogen release into EAST discharges was identified as the graphite tiles used in the lower divertor. The hydrogen content in EAST plasmas were clearly reduced by first employing a high-temperature vacuum baking of all graphite tiles and then renewing a 100–200 μm thick SiC coating before an EAST experimental run campaign. Subsequent active surface conditioning of all wall components with elemental silicon and then with elemental lithium were seem to again reduce the plasma hydrogen content significantly – with lithium proving to be more effective than silicon. Combining these several techniques, H/(H + D) levels as low as ~3% have been achieved in EAST discharges. Additionally, the effects of lithium thickness on H surface implantation and retention has been re-examined semi-quantitatively using data from a previous run campaign. These data suggest that relatively thick Li films coated on the first wall can effectively isolate the rich source of hydrogen stored in the porous bulk of the underlying graphite from the deuterium-fueled plasma so as to minimize hydrogen release. In conclusion, an operational maneuver whereby a diverted plasma is repetitively switched from an upper single null configuration to a lower single null configuration is presented. This switching maneuver has been shown to suppress hydrogen influx into a 35s-long EAST discharge by alternately mitigating the rise in divertor temperature.

[en] Water has both advantages and disadvantages as a coolant in conceptual designs of future fusion power plants. In the United States, water has not been chosen as a fusion power core coolant for decades. Researchers in other countries continue to adopt water in their designs, in some cases as the leading or sole candidate. In this article, we summarize the technical challenges resulting from the choice of water coolant and the differences in approach and assumptions that lead to different design decisions amongst researchers in this field.

[en] The modeling capability for tubes with twisted tape inserts is reviewed with reference to the application of cooling plasma facing components in magnetic confinement fusion devices. The history of experiments examining the cooling performance of tubes with twisted tape inserts is reviewed with emphasis on the manner of heating, flow stability limits and the details of the test section and fluid delivery system. Models for heat transfer, burnout, and onset of net vapor generation in straight tube flows and tube with twisted tape are compared. As a result, the gaps in knowledge required to establish performance limits of the plasma facing components are identified and attributes of an experiment to close those gaps are presented

[en] In this study, a conceptual design for a pre-filled liquid lithium divertor target for the National Spherical Torus Experiment Upgrade (NSTX-U) is presented. The design is aimed at facilitating experiments with high lithium flux from the plasma facing components (PFCs) in NSTX-U and investigating the potential of capillary based liquid lithium components. In the design, lithium is supplied from a reservoir in the PFC to the plasma facing surface via capillary action in a wicking structure. This working principle is also demonstrated experimentally. Next, a titanium zirconium molybdenum (TZM) prototype design is presented, required to withstand a steady state heat flux peaking at 10 MW m"–"2 for 5 s and edge localized modes depositing (130 kJ in 2 ms at 10 Hz). The main challenge is to sufficiently reduce the thermal stresses. This is achieved by dividing the surface into brushes and filling the slots in between with liquid lithium. The principle of using this liquid “interlayer” allows for thermal expansion and simultaneously heat conduction, and could be used to significantly reduce the demands to solids in future PFCs. Lithium flow to the surface is analyzed using a novel analytical model, ideally suited for design purposes. Thermal stresses in the PFC are analyzed using the finite element method. As a result, the requirements are met, and thus a prototype will be manufactured for physical testing.

[en] The superconducting magnet systems of future fusion reactors, such as a Demonstration Power Plant (DEMO), will produce magnetic field energies in the 10 s of GJ range. The release of this energy during a fault condition could produce arcs that can damage the magnets of these systems. The public safety consequences of such events must be explored for a DEMO reactor because the magnets are located near the DEMO's primary radioactive confinement barrier, the reactor's vacuum vessel (VV). Great care will be taken in the design of DEMO's magnet systems to detect and provide a rapid field energy dump to avoid any accidents conditions. During an event when a fault condition proceeds undetected, the potential of producing melting of the magnet exists. If molten material from the magnet impinges on the walls of the VV, these walls could fail, resulting in a pathway for release of radioactive material from the VV. A model is under development at Idaho National Laboratory (INL) called MAGARC to investigate the consequences of this accident in a large toroidal field (TF) coil. Recent improvements to this model are described in this paper, along with predictions for a DEMO relevant event in a toroidal field magnet

[en] Active control of the toroidal current density profile is among those plasma control milestones that the National Spherical Torus eXperiment-Upgrade (NSTX-U) program must achieve to realize its next-step operational goals. Motivated by the coupled, nonlinear, multivariable, distributed-parameter plasma dynamics, the first step towards control design is the development of a physics-based, control-oriented model for the current profile evolution in response to non-inductive current drives and heating systems. The evolution of the toroidal current density profile is closely related to the evolution of the poloidal magnetic flux profile, whose dynamics is modeled by a nonlinear partial differential equation (PDE) referred to as the magnetic-flux diffusion equation (MDE). The proposed control-oriented model predicts the spatial-temporal evolution of the current density profile by combining the nonlinear MDE with physics-based correlations obtained at NSTX-U for the electron density, electron temperature, and non-inductive current drives (neutral beams). The resulting first-principles-driven, control-oriented model is tailored for NSTX-U based on predictions by the time-dependent transport code TRANSP. Furthermore, main objectives and possible challenges associated with the use of the developed model for the design of both feedforward and feedback controllers are also discussed.

[en] Additive Manufacturing (AM) can create novel and complex engineered material structures. Features such as controlled porosity, micro-fibers and/or nano-particles, transitions in materials and integral robust coatings can be important in developing solutions for fusion subcomponents. A realistic understanding of this capability would be particularly valuable in identifying development paths. Major concerns for using AM processes with lasers or electron beams that melt powder to make refractory parts are the power required and residual stresses arising in fabrication. A related issue is the required combination of lasers or e-beams to continue heating of deposited material (to reduce stresses) and to deposit new material at a reasonable built rate while providing adequate surface finish and resolution for meso-scale features. In conclusion, Some Direct Write processes that can make suitable preforms and be cured to an acceptable density may offer another approach for PFCs.