Plasma-wall interaction plays an essential role in the performance and safety of a fusion reactor. This thesis focuses on the impact of material migration on plasma-facing components. It is based on experiments performed in tokamaks: JET, TEXTOR and ASDEX Upgrade. The objectives of the experiments were to assess fuel and impurity removal under ion cyclotron wall conditioning (ICWC) and plasma impact on diagnostic mirrors. In wall conditioning studies, tracer techniques based on the injection of rare isotopes (15N, 18O) were used to determine conclusively the impact of the respective gases. For the first time, probe surfaces and wall components exposed to ICWC were examined by surface analysis methods. Discharges in hydrogen were the most efficient to erode carbon co-deposits, resulting in a reduction of the initial deuterium content by a factor of two. It was also found that impurities desorbed under ICWC are partly re-deposited on the wall. Plasma impact on diagnostic mirrors was determined by surface analysis of test mirrors exposed at JET. Reflectivity of mirrors from the divertor region was severely decreased due to deposits of beryllium, deuterium, carbon and other impurities. This result points out the need to develop mirror maintenance procedures. Neutron damage on mirrors was simulated by ion irradiation in an ion implanter. It was shown that damage levels similar to those expected in the first wall of a fusion reactor do not produce a significant change in reflectivity. / <p>QC 20160819</p>

Fuel inventory and generation of carbon and metal dust in a tokamak are perceived to be serious safety and economy issues for the steady-state operation of a fusion reactor, e.g. ITER. These topics have been explored in this thesis in order to contribute to a better understanding and the development of methods for controlling and curtailing fuel accumulation and dust formation in controlled fusion devices. The work was carried out with material facing fusion plasmas in three tokamaks: TEXTOR in Forschungszentrum Jülich (Germany), Tore Supra in the Nuclear Research Center Cadarache (France) and JET in Culham Centre for Fusion Energy (United Kingdom). Following issues were addressed: (a) properties of material migration products, i.e. co-deposited layers and dust particles; (b) impact of fuel removal methods on dust generation and on modification of plasma-facing components; (c) efficiency of fuel and deposit removal techniques; (d) degradation mechanism of diagnostic components - mirrors - and methods of their regeneration. / <p>QC 20121116</p>

The promise of a vast source of energy to power the world and protect our planet using fusion technology has been the driving force for scientists and engineers around the globe for more than sixty years. Although the materialization of this ideal still in the distance, multiple scientific and technological advances have been accomplished, which have brought commercial fusion power closer to a reality than it has ever been. As part of the collaborative effort in the pursuit of realizable fusion energy, the International Thermonuclear Experimental Reactor (ITER) is being developed by a coalition of nations of which the United States is a part of. One critical technological challenge for ITER is the development of adequate plasma facing materials (PFMs) that can withstand the strenuous conditions of operation. To date, high heat flux (HHF) testing has been conducted mainly on non-irradiated specimens due to the difficulty of working with radioactive specimens, such as instrument contamination. In this thesis, the new Irradiated Material Target Station (IMTS) facility for fusion materials at Oak Ridge National Laboratory (ORNL), in which the HHFs are provided by water-wall plasma-arc lamps (PALs), is considered for neutron-irradiated specimens, especially tungsten. The facility is being used to test irradiated plasma-facing components materials for magnetic fusion reactors as part of the US-Japan plasma facing components evaluation by tritium plasma, heat and neutron irradiation experiments (PHENIX). In order to conduct HHF testing on the PFMs various sample holders designs were developed to accommodate radioactive specimens during HHF testing. As part of the effort to design sample holders that are compatible with the IMTS facility, numerical simulations were performed for different water-cooled sample holder designs with the commercial computational fluid dynamics (CFD) software package, ANSYS™ FLUENT®. The numerical models are validated against experimental temperature measurements obtained from the IMTS facility. These experimentally validated numerical models are used to assess the thermal performance of two sample holder designs and establish safe limits for HHF testing under various operating conditions. The limiting parameter for the current configuration was determined for each sample holder design. For the Gen 1 sample holder, the maximum temperature reached within the Copper rod limits the allowable incident heat flux to about 6 MW/m². In the case of the Gen 2 sample holder, the maximum temperature reached within the Molybdenum clamping disk limits the allowable incident heat flux to about 5 MW/m². In addition, the numerical model are used to parametrically investigate the effect of the operating pressure, mass flow rate, and incident heat flux on the local heat flux distributions and peak surface temperatures. Finally, a comparative analysis is conducted to evaluate the advantages and disadvantages associated with the main design modifications between the two sample holder models as to evaluate their impact in the overall thermal performance of each sample holder in order to provide conclusive recommendations for future sample holder designs.

Material erosion, melting, splashing, bubbling, and ejection during disruption events in future large tokamak reactors are of serious concern to component longevity. The majority of the heat flux during disruptions will be incident on the divertor, which will be made from tungsten in the future large tokamak ITER. Electrothermal plasma sources operating in the confined controlled arc discharge regime produce heat fluxes in the range expected for hard disruptions in future large tokamaks. The radiative heat flux produced inside of the capillary discharge channel is from the formed high density (10^23 - 10^27/m^3) plasma with heat fluxes of up to 125 GW/m^2 over a period of 100s of microseconds, making such sources excellent simulators for ablation studies of plasma-facing materials in tokamaks during hard disruptions. Experiments have been carried out with the PIPE device exposing tungsten to these high heat flux plasmas. SEM images have been taken of the tungsten surfaces, cross sections of tungsten surfaces, and ejected material. Melting and bubble/void formation has been observed on the tungsten surface. The tungsten surface shows evidence of melt-layer flow and the existence of voids and cracks in the exposed material. The ejected material does not show direct evidence of liquid material ejection which would lead to splashing. EDS analysis has been performed on the ejected material which demonstrates a lack of deposited solid tungsten particulates greater than micron size. / Master of Science

Tungsten (W) based materials can be used in fusion reactors due to several advantages. Different fabrication routes can be applied to develop tungsten materials with intended microstructure and properties for specific application including nanostructured grades. Therein, innovative chemical routes are unique in their approach owing numerous benefits. This thesis summarizes the development of W-based composites dispersed-strengthened by rare earth (RE) oxides and their evaluation for potential application as plasma facing armour material to be used in fusion reactor. Final material development was carried out in two steps; a) fabrication of nanostructured metallic tungsten powder dispersed with RE-oxides and b) powder sintering into bulk oxide-dispersed strengthened (ODS) composite by spark plasma process. With the help of advanced characterization tools applied at intermediate and final stages of the material development, powder fabrication and sintering conditions were optimized. The aim was to achieve a final material with a homogenous fine microstructure and improved properties, which can withstand under extreme conditions of high temperature plasma. Two groups of starting materials, synthesized via novel chemical methods, having different compositions were investigated. In the first group, APT-based powders doped with La or Y elements in similar ways, had identical particles’ morphology (up to 70 μm). The powders were processed into nanostructured composite powders under different reducing conditions and were characterized to investigate the effects on powder morphology and composition. The properties of sintered tungsten materials were improved with dispersion of La2O3 and Y2O3 in the respective order. The oxide dispersion was less homogeneous due to the fact that La or Y was not doped into APT particles. The second group, Ydoped tungstic acid-based powders synthesized through entirely different chemistry, contained nanocrystalline particles and highly uniform morphology. Hydrogen reduction of doped-tungstic acid compounds is complex, affecting the morphology and composition of the final powder. Hence, processing conditions are presented here which enable the separation of Y2O3 phase from Y-doped tungstic acid. Nevertheless, the oxide dispersion reduces the sinterability of tungsten powders, the fabricated nanostructured W-Y2O3 powders were sinterable into ultrafine ODS composites at temperatures as low as 1100 °C with highly homogeneous nano-oxide dispersion at W grain boundaries as well as inside the grain. The SPS parameters were investigated to achieve higher density with optimum finer microstructure and higher hardness. The elastic and fracture properties of the developed ODS-W have been investigated by micro-mechanical testing to estimate the materials’ mechanical response with respect to varying density and grain size. In contrast from some literature results, coarse grained ODS-W material demonstrated better properties. The developed ODS material with 1.2 Y2O3 dispersion were finally subjected to high heat flux tests in the electron beam facility “JUDITH-1”. The samples were loaded under ELM-like thermal-shocks at varying base temperatures up to an absorbed power density of 1.13 GW/m2, for armour material evaluation. Post mortem characterizations and comparison with other reference W grades, suggest lowering the oxide contents below 0.3 wt. % Y2O3. As an overview of the study conducted, it can be concluded that innovative chemical routes can be potential replacement to produce tungsten based materials of various composition and microstructure, for fusion reactor applications. The methods being cheap and reproducible, are also easy to handle for large production at industrial scale. / <p>QC 20120827</p>

科学研究費補助金 研究種目:一般研究(A) 課題番号:01420044 研究代表者:高村 秀一 研究期間:1989-1991年度

Thermonuclear fusion may become an attractive future power source. The most promising of all fusion machine concepts is the tokamak. Despite decades of active research, still huge tasks remain before a fusion power plant can go online. One of these important tasks deals with the interaction between the fusion plasma and the reactor wall. This work focuses on how eroded wall materials of different origin and mass are transported in a tokamak device. Element transport can be examined by injection of certain species of unique and predetermined origin, so called tracers. Tracer experiments were conducted at the TEXTOR tokamak before its final shutdown. This offered an unique opportunity for studies of the wall and other internal components: For the first time it was possible to completely dismantle such a machine and analyse every single part of reactor wall, obtaining a detailed pattern of material migration. Main focus of this work is on the high-Z metals tungsten and molybdenum, which were introduced by WF6 and MoF6 injection into the TEXTOR tokamak in several material migration experiments. It is shown that Mo and W migrate in a similar way around the tokamak and that Mo can be used as tracer for W transport. It is further shown how other materials - medium-Z (Ni), low-Z (N-15 and F), fuel species (D) - migrate and get deposited. Finally, the outcome of dust sampling studies is discussed. It is shown that dust appearance and composition depends on origin, formation conditions and that it can originate even from remote systems like the NBI system. Furthermore, metal splashes and droplets have been found, some of them clearly indicating boiling processes. / <p>QC 20151203</p>