Conference Report on the 7th International Symposium on Liquid metals Applications for fusion (ISLA-7)

Supported by the world magnetic fusion research community, a series of International Symposia on Liquid metals Applications for fusion (ISLA) have been held biannually since 2010. The 7th edition (ISLA-7) was held for the period from 12 December through 16 December 2022, at Chubu University located in Kasugai, Aichi, Japan. For the first time in the history of this series of symposia, ISLA-7 was held in a hybrid fashion, due to the COVID-19 situation. The total number of the participants was 60, 34 out of whom attended the symposium in person, and the rest participated online. As to the presentation statistics, 29 papers were presented in person, whereas 21 presentations were delivered online but real-time by the presenters in China, Spain, the UK, and the USA. Both of the presentations delivered in person and online were recorded, and the video has been shared by all participants. These participants represent 11 countries: China, Czech, Italy, Japan, Latvia, Netherlands, Russia, Thailand, the UK, and the USA. All these numbers are among the largest in this series of symposia. Covered by these presentations are; in session-2, program overviews and liquid metal research review; in session-3, liquid metal flows, and MHD issues; in session-4, liquid metal facilities; in sessions-5 and 6, liquid metal experiments and modeling; in session-7, divertor physics and heat flux mitigation; in session-8, plasma and liquid metals interactions; in session-9 liquid metal plasma-facing components, erosion, and wettability. In addition, there were an opening session whereby several opening addresses were delivered and also a closing session where all technical session summaries were presented by the respective session chairs.journal articl

The design and development of hydrogen isotope extraction technologies for a limit-style liquid lithium loop

As lithium has grown in popularity as a plasma-facing material, efforts have been placed on examining its viability as a first wall candidate. Lithium has proven over numerous studies to improve core confinement, while allowing access to operational regimes previously unattainable while using solid, high-Z divertor and limiter modules. These benefits are due to the fuel retention capabilities of lithium, which allow it to be an almost ideally absorbing boundary, which is both beneficial and problematic. While lithium exhibits a number of other advantages and disadvantages as a plasma-facing material, none is more important than the tritium retention problem. As such, extraction technologies must be constructed and verified within the scope of larger scale lithium loop systems that separate lithium impurities, recover deuterium and tritium, and recycle clean liquid lithium back to the plasma-material interface. Laboratory-scale and pilot-scale studies have been conducted at the Center for Plasma-Material Interactions at the University of Illinois to investigate a number of phenomena that influence the recovery of entrained tritium from lithium. While the ultimate goal is to develop a fully-functional liquid lithium loop for the Lithium Metal Infused Trenches plasma-facing component, complete with efficient hydrogen reclamation technologies, there exists a lack in understanding within the community of the thermochemical fundamentals that are envisioned to drive tritium reclamation. Of specific interest are the evolution fluxes of hydrogen isotopes from solutions of various concentrations of hydrogen in lithium, and the associated temperatures. The knowledge of how the isotopic fraction affects recovery is pivotal to determining the appropriate thermal treatment technique. The laboratory-scale experiments in this report aimed at filling in the knowledge gaps in the literature with regards to the thermochemistry of the hydrogen-lithium system. In all cases, hydrogen was used as an isotopic surrogate for deuterium and tritium. Success was based on an individual samples ability to evolve molecular hydrogen at rates that would match or exceed in-vessel wall losses, determined from a simulated Lithium-Walled International Thermonuclear Experimental Reactor scenario. The hydrogen degassing of pure lithium hydride was observed to exceed fuel loss by a factor of two or greater, at temperatures near the melting point for hydride. Samples of both solid and liquid lithium were subjected to different hydrogen environments under a variety of exposure conditions. During plasma exposures, evidence of saturation, where hydride layers are formed at or near the sample surface and inhibit hydrogen absorption, was witnessed for solid lithium samples. Liquid samples exhibited this behavior to a lesser degree; however, mass diffusion was able to transport the insulating species away from the surface and absorption was able to continue, albeit to a lesser extent than was initially detected. The sub-surface chemistry was found to still be limited by the thermodynamic solubility thresholds in a plasma environment, meaning enhanced hydrogen dissolution was not witnessed at ion energies relevant to these experiments. The presence of a plasma, however, did appear to enhance absorption rates above and beyond what was capable with hydrogen gas alone. During these tests, hydrogen evolution rates from the dissolved phase never approached the point of being able to balance losses at the plasma-material interface, being always less by a factor of two or more. It was therefore determined that supplementary methods were required to enhance thermal-based recovery in solutions with hydrogen molar ratios less than the solubility limits. This work culminated in the design, development, construction, and proof-of-concept testing of a distillation column. Envisioned to be an integrated treatment method in a fully functional lithium loop, the column was developed based on the need to recover tritium and recycle fresh lithium back into the reactor. The novelty in this design was in its use of induction heating drive and condensation stages. Proof-of-concept tests were performed in the fully constructed prototype with solutions of lithium and lithium hydride at various molar ratios. The system was observed to operate as intended during these initial runs, but requires further testing; however, the column marks the first system constructed for the sole purpose of recovering tritium from a lithium-walled reactor. Such a system will prove most effective if upstream separation and purification techniques are present to divert the lithium deuteride and lithium tritide-rich streams to the column for thermal decomposition and degassing. In the case where upstream purification modules are absent from the lithium loop, the column alone will be hard pressed to achieve recovery rates in far-from-saturated solutions that balance wall losses. A technique to supplement the induction heating drive was therefore proposed. Ultrasonic degassing of liquid metals is an industry-tested technique used to rid melts of dissolved gases by taking advantage of acoustically-induced cavitation. This process was theoretically applied to the hydrogen-lithium system, displaying evidence that degassing is most effective in the presence of heat, ultrasonic waves, and vacuum. This work laid the theoretical groundwork for future application. The results presented in this report show that using the appropriate combination of treatment methods, hydrogen, and by extension deuterium and tritum, can be recovered from lithium at rates that balance in-vessel wall loss. Future work will be needed to then integrate these methods into a fully functional liquid lithium loop

Phenomenology of plasma-wall interaction using liquid metals in tokamak devices

The goal of the present thesis is to present the theoretical and experimental results concerning the use of liquid metal in tokamak devices. Most of the experimental work has been performed on FTU at the ENEA laboratories in Frascati. The main part of the work has consisted of analyzing the evolution of thermal loads, plasma contamination and plasma edge parameters variations determined by the exposure of liquid metal limiters. A radiative model is proposed to explain the vapor shield effect observed during the past experimental campaigns with the Liquid Lithium Limiter: the results of the simulations and the comparisons made with the experimental data are illustrated. Much time has been spent in the installation, debugging and exploration of two different limiters on FTU: the Cooled Lithium Limiter and the Liquid Tin Limiter. FTU was the first, and up to now unique, tokamak in the world operating with a liquid tin limiter. The results, supported by experimental data and simulations, are reported and discussed. Moreover, a comparison between lithium and tin experiments on FTU is presented. Part of the Ph.D. work has been dedicated to the study and the interpretation of the diagnostics deputy to the main plasma edge parameters measurements. Different Langmuir probes were designed and built for experiments in FTU and other devices too. The old FTU Langmuir probes acquisition system has been successfully moved to a new hardware, and a new software is now performing the data reconstruction. Another part of the work has concerned the implementation ex novo of a dedicated laboratory in Frascati focused on the liquid metal features studies, such as wettability, corrosion, and chemical-physical characteristics investigation of the fusion relevant liquid metals (lithium, gallium, and tin). A high vacuum oven was installed for this purpose, and the first test phase ended successfully. Furthermore, a small vacuum chamber was assembled allowing temperature up to 1500°C in a limited volume. This made it possible to increase the testing rate and ultimately to achieve the wetting of small-size tungsten CPS structures with both gallium and tin. In the present work an innovative tungsten coating processes, to avoid corrosion of the structural materials has been investigated. The studies conducted together with the University ”La Sapienza” and the ENEA Brasimone research center led to the validation and the realization of a solid and reliable tungsten deposit using the detonation gun machine. During the final period of the Ph.D., liquid tin samples analyses have been carried out after the plasma exposure on ISTTOK tokamak in Lisbon. The results obtained in the heat load calculation and tin effects on plasma are presented, together with a new proposal for the use of CPS-base samples. Finally, a work undertaken in collaboration with the INFN of Frascati and the CERN laboratory in Geneva is reported, regarding the possibility of using liquid lithium as a target for a particle accelerator for muon production

Hydrogenic and impurity retention studies on liquid lithium and tungsten as materials for a nuclear fusion reactor by glow discharge and laser techniques

Tesis de la Universidad Complutense de Madrid, Facultad de Ciencias Físicas, Departamento de Física Atómica, Molecular y Nuclear, leída el 12-01-2018La energía de fusión es un proceso perfectamente viable desde el punto de vista científico. El Sol y las demás estrellas del universo actúan como “reactores de fusión” que funcionan adecuadamente día tras día. Un ejemplo de ello lo tenemos en nuestro Sol cuya superficie emite una enorme cantidad de radiación (~4.5∙1020 W), consumiendo para ello unas 5 toneladas de hidrógeno cada segundo. Sin embargo la reproducción de este proceso en nuestro planeta a través de dispositivos controlados de fusión magnética que puedan producir electricidad resulta extremadamente complicada debido a desafíos tecnológicos muy importantes. Entre ellos, la selección de materiales en contacto con el plasma, capaces de extraer la potencia generada y resistir bajo las extremas condiciones esperadas en el interior de estos reactores, es uno de los asuntos más críticos a resolver. El tungsteno y el litio líquido están entre los candidatos mejor considerados para conseguir este objetivo. Esta tesis explora la utilización de estos elementos para conformar estos componentes, enfatizando en dos importantes problemas derivados de su uso: la formación de amoníaco (tritiado) durante las descargas con “seeding” de N2 y la potencial absorción hidrogénica (tritio) en capas híbridas litio líquido-tungsteno...Fusion energy is a perfectly feasible process from the scientific point of view. The Sun and the rest of stars act as "fusion reactors" that work properly day by day. A clear example can be found in the Sun, whose surface emits a huge radiation power (~4.5∙1020 W), employing for this purpose a hydrogen amount of 5 tons per second approximately. However, the reproduction of this process in our planet by means of magnetic controlled fusion devices that could produce electric power results extremely complicated due to very important technological challenges. Among them, the selection of the materials in contact with the plasma, able to extract the generated power and resist under the extreme conditions expected in such reactors, is one of the most critical issues to solve. Tungsten and liquid lithium are among the candidates better considered to develop this purpose. This thesis explores the utilization of these elements to conform the “plasma facing components”, emphasizing in two important problems derived of their use: the formation of (tritiated) ammonia during N2 seeded discharges and the potential hydrogenic (tritium) uptake in tungsten-liquid lithium mixed layers...Depto. de Estructura de la Materia, Física Térmica y ElectrónicaFac. de Ciencias FísicasTRUEunpu

Nuclear Fusion Programme: Annual Report of the Association Karlsruhe Institute of Technology (KIT)/EURATOM ; January 2009 - December 2009 (KIT Scientific Reports ; 7548)

The Karlsruhe Institute of Technology (KIT) is working in the framework of the European Fusion Programme on key technologies in the areas of superconducting magnets, microwave heating systems (Electron-Cyclotron-Resonance-Heating, ECRH), the deuterium-tritium fuel cycle, He-cooled breeding blankets, a He-cooled divertor and structural materials, as well as refractory metals for high heat flux applications including a major participation in the preparation of the international IFMIF project

Nuclear Fusion Programme: Annual Report of the Association Karlsruhe Institute of Technology/EURATOM ; January 2011 - December 2011 (KIT Scientific Reports ; 7621)

The Karlsruhe Institute of Technology (KIT) is working in the framework of the European Fusion Programme on key technologies in the areas of superconducting magnets, microwave heating systems (Electron-Cyclotron-Resonance-Heating, ECRH), the deuterium-tritium fuel cycle, He-cooled breeding blankets, a He-cooled divertor and structural materials, as well as refractory metals for high heat flux applications including a major participation in the preparation of the international IFMIF project