May 12th, 2022
MHD modelling for liquid metal systems and components, by Alessandro Tassone
Abstract: Liquid metals are promising fluids for applications in breeding blankets (BB) and plasma facing components (PFC) but, owing to their high electrical conductivity, tend to behave in bizarre and counter-intuitive ways when exposed to the intense magnetic fields which are typical of magnetic confinement reactors. Comprehension and characterization of the magnetohydrodynamic (MHD) phenomena is necessary to successfully develop and deploy components based on liquid metal technology. In this contribution, the most relevant effects of MHD for BB and PFC are reviewed and the work done at Sapienza University of Rome to model these phenomena is presented. The focus will be on direct numerical simulations performed with computational fluid-dynamic codes and the establishment of a framework for a system level code to be used in the future for safety analyses.
Zoom link password: C8C39027
May 5, 2022
Divertor heat load estimation with thermal measurements in WEST, by Jonathan Gaspar
Abstract: WEST is a full W tokamak with an extensive set of diagnostics for heat load measurements especially in the lower divertor. It is composed by infrared thermography, thermal measurement with thermocouples and fiber Bragg grating embedded few mm below the surface. Associated to this thermal measurement different inverse method have been developed to estimate the heat flux deposited on the plasma facing components. The plasma heat flux is characterized by the time evolution of its amplitude and spatial shape on the target (heat flux decay length ðœ†ð‘žð‘¡, power spreading in the private flux region ð‘†ð‘¡ and the strike point location ð‘¥0). The thermal diagnostics and the inverse method will be described and illustrated with experimental data from different campaigns performed at WEST.
Zoom link password: 80565B4B
March 24, 2022
Developing Tungsten-diamond composites for fusion applications, by Anneqa Khan
Abstract: The challenges for plasma facing materials in a fusion reactor are well documented. This talk will explore the scope for developing a new tungsten-diamond composite that could help address these challenges..
Zoom link password: 4E9386C5
March 15, 2022
Understanding the plasma with integrated modelling, by Michele Marin
Abstract: Accurately reproducing the plasma dynamics requires complex simulations, which can take considerable computing resources. Reduced models can greatly speed up the process, but this comes at the cost of assumptions and simplifications. Therefore, these faster models need to be carefully validated and compared with other codes and with the experiments. Integrated modelling is a technique that evolves a number of the tokamak subsystems at the same time, improving consistency and easing the comparison with experiments. The talk includes integrated modelling, its validation cycle and examples of applications..
Zoom link password: 9B1E84B6
February 22nd, 2022
Plasma cleaning of diagnostic mirrors in ITER, by Kunal Soni (FUSION-EP 2015)
Abstract: Nearly 40 optical diagnostic systems in ITER are equipped with metallic first mirrors (FMs) with the objective of directing the light from the fusion plasma towards the diagnostics through an optical labyrinth in order to prevent neutron leakage. However, the FMs being the initial elements in the optical diagnostics, would be subject to constant erosion from charge exchange neutrals as well as deposition of the first wall materials: beryllium (Be), tungsten (W) and their oxides, that would significantly degrade their optical properties. The FMs would hence require regular cleaning to restore their optical properties, currently foreseen to be achieved by an in-situ capacitively coupled radio-frequency (CCRF) plasma cleaning technique. In this contribution we discuss the plasma cleaning of such FMs in ITER relevant conditions..
Zoom link password: 7D8D1AE5
February 16th, 2022
Advanced divertors. A cutting edge tool to control plasma exhaust, by Cyd Cowley
Abstract: The idea behind magnetic confinement fusion is to create a well-confined star-like plasma on Earth. Eventually, however, the fusion plasma in our devices must touch the surrounding material, which can pose significant challenges given that plasma can reach temperatures up to 10 times hotter than the core of the sun. In an attempt to reduce and control this plasma exhaust, advanced divertors have been proposed, which implement novel magnetic or geometric topologies of the plasma. Taking advantage of divertor and detachment physics, advanced divertors and their features may be crucial for safe operation of next-generation tokamaks.
Zoom link password: 3D1300A3
February 09th, 2022
Masterclass: Advanced materials and new concepts for future fusion devices, by Andrey Litnovsky
Abstract: The construction of the fusion power station, such as DEMOnstration power plant (DEMO) represents an ultimate scientific, engineering and safety challenge. In particular, materials facing fusion plasmas will have to withstand simultaneous neutron and plasma exposures and heat loads in the course of several hours of operation during a single plasma discharge of the power plant. These conditions will challenge material properties severely. It is clear already now, that conventional plasma-facing materials used in present-day fusion devices as well as in future fusion experiments, may not fully satisfy the requirements of a safe and reliable operation. The lecture will provide a â€œglimpseâ€ into the creation of advanced materials for fusion. Based on a vast material science knowledge supported by new manufacturing technologies, it was possible to overcome several crucial natural limitations of conventional materials, such as tungsten both in terms of plasma performance as well as of their safety. The advanced tungsten-fiber-reinforced tungsten composites as well as micro-structured tungsten materials are being developed for a quasi steady-state operation under heat loads reaching 20 MW/m2 in the divertor of the power plant. The micro-structured tungsten demonstrated the unbelievable resilience and no detectable damage after the transient heat loads of up to 0.64 GW/m2 for 200.000 cycles. The innovative self-passivating SMART alloys ensure an acceptable plasma performance of the first wall of DEMO while maintaining the unprecedented 105-fold suppression of natural tungsten oxidation and more than 40-fold suppression of sublimation of tungsten oxide in case of an accident at the fusion power plant. The spectacular tungsten laminates feature the material ductility already at room temperature. These and other remarkable material concepts will be reviewed along with advanced manufacture techniques, such as additive manufacturing and field-assisted sintering technology. These technologies are used for scaling up the advanced materials to the reactor-relevant sizes thus making a crucial step toward the final goal â€“ building of first wall and divertor components of the fusion power plant.
Zoom link password: 9617ECE8
January 25th, 2022
SOLPS-ITER studies of Neon seeding in EAST, by Dieter Boeyaert
Abstract: In order to decrease power and particle loads towards the divertor targets in future fusion devices, active extrinsic impurity seeding is required . A dedicated experimental program on neon seeding in H-mode plasmas was performed at EAST . Experimental observations show a factor of three reduction of the power flux towards the targets, but this is not sufficient to determine whether detachment is reached. Therefore, the SOLPS-ITER code package  is employed for assessing the plasma edge transport in the EAST discharges of ref. . Drift terms are successfully included, resulting in upstream agreement within the experimental error bars and downstream agreement within a factor three or better between simulations and experiments, and showing that Ne seeding induces detachment. Sensitivity studies towards the main unknown input parameters for the code are executed. In order to quantify the precision of the performed simulations, the numerical errors affecting the simulation results are examined. Similar to previous ITER simulations with B2-EIRENE , the main error is the discretization error due to the finite plasma grid. By making an appropriate choice of the remaining numerical input parameters, the error contributions induced by the Monte Carlo noise of the EIRENE code are kept sufficiently small.  M. Wischmeier et al., J. Nucl. Mater. 463 (2015) 22-29  D. Boeyaert et al., Nucl. Mater. Energy 26 (2021) 100926  S. Wiesen et al., J. Nucl. Mater. 463 (2015) 480-484  K. Ghoos et al., Nucl. Fusion 59 (2018) 026001.
Zoom link: fusion.yt/bf
January 18th, 2022
Retention and recycling of D/T fuels in a fusion reactor, by Pr. Tetsuo Tanabe (Osaka City University)