The heat transfer performance of the Helium-cooled Multi-jet (HEMJ) divertor was investigated. The HEMJ design uses impinging jets to significantly enhance its heat transfer capability. The convective heat transfer coefficient predicted by computational fluid dynamics software packages is on the order of 50,000 W/(m2-K). The high predicted values of the convective heat transfer coefficient necessitated experimental validation, which was the focus of this investigation. A test section which simulates the thermal performance of the HEMJ divertor was designed, constructed, and instrumented for testing an in air flow loop. The operating conditions of the air flow loop were chosen to match the non-dimensional operating conditions expected for the HEMJ divertor in a post-ITER fusion power plant. The air flow loop experiments were performed for mass flow rates of 2.0 g/s to 8.0 g/s and with incident nominal heat fluxes of 0.8 MW/m2 and 1.0 MW/m2. The angular variation of the heat transfer coefficient was also investigated. Numerical simulations which matched the experimental operating conditions were performed using the computational fluid dynamics software package, FLUENT® 6.2. Comparisons of the experimental and numerical pressure drop, temperature, and heat transfer coefficient were made. The experimental results agreed with the numerical predictions for all operating conditions in this investigation. This provided a strong degree of confidence in using the FLUENT® software package to analyze the HEMJ divertor design.

Sustainable fusion power is within reach; however, more research is needed in the field of material science and engineering. One critical component of a fusion reactor is the plasma facing material. Very little literature exists on the sustainability of tungsten as a plasma facing material (PFM). During operation, PFM must withstand harsh conditions with combined effects from high temperature, mechanical stress, irradiation, transmutation, and the production of hydrogen (H) and helium (He) from nuclear reactions. Therefore, this thesis will focus on co-implantation of H and He into tungsten to investigate the mechanical and microstructural material response. For the first part of this study, Molecular Dynamics (MD) was used to qualitatively understand defect migration and mechanical property changes in tungsten. A Brinell hardness test was simulated using MD in tungsten to study the dependence on void size and void density hardness. It was found that hardness changes vary as the square root of the void size and void density. Also the movement of dislocations and its interaction with voids were investigated. For the second part of the study, H and He were co-implanted into tungsten to look at the mechanical and microstructural changes. Hardness changes were measured using a nano-indenter ex-situ on post-irradiated specimen. Results show that the hardness of tungsten after co-implantation is proportional to the square root of the fluence. Additionally, the microstructure of irradiated tungsten samples was investigated by using a Transmission Electron Microscope (TEM). It was observed that the defect microstructure in tungsten, after co-implantation, is quite complex, with a number of intriguing features, such as the presence of the nano-bubbles and dislocation loops. Also it was observed that there was an effect that H has on the nucleation of He nano-bubbles. The results from this work suggest that the effect of co-implanting H and He into tungsten is crucial to fully understand its viability as a PFM.

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>

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

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>

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.

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>

Experimental and numerical studies simulating the gas-cooled divertor plate design concept have been carried out. While thermo-fluid and thermo-mechanical analyses have been previously performed to show the feasibility of the divertor plate design and its ability to accommodate a maximum heat flux of up to 10 MW/m2, no experimental data have heretofore been published to support or validate such analyses. To that end, this investigation has been undertaken. A test module with prototypical cross-sectional geometry has been designed, constructed, and instrumented. Experiments spanning the prototypical Reynolds numbers of the helium-cooled divertor have been conducted using pressurized air as the coolant. A second test module where the planar jet exiting the inlet manifold is replaced by a two-dimensional hexagonal array of circular jets over the entire top surface of the inlet manifold has also been tested. The thermal performance of both test modules with and without a porous metallic foam layer in the gap between the outer surface of the inlet manifold and the cooled surfaces of the pressure boundary were directly compared. For a given mass flow rate, the slot design with the metallic foam insert showed the highest heat transfer coefficient, with a pressure drop lower than that of the array of circular jets without foam. Additionally, numerical simulations matching the experimental operating conditions for the two cases without foam were performed using the computational fluid dynamics software package, FLUENT® v6.2. Comparisons of the experimental and numerical pressure drop, temperature, and heat transfer coefficient were made.

Fusion energy is a potential candidate for sustain-able steady-state energy supply. However, a fully functional fusion reactor is not yet available and several remaining challenges need to be addressed before fusion becomes a reliable source. One of the remaining challenges with fusion is the plasma-induced modification of the inner wall of the tokamak, i.e. the structures surrounding hot plasma. Due to the rarity of tritium, an important element in future fusion fuel, the plasma facing component (PFC) should have as low fuel retention as possible. In this thesis, methods for controlling ion accumulation in a material sample have been developed. Using the new system, a molybdenum (42Mo) target has been implanted with deuterium (2H) and the retention has been measured with ion beam analysis. The experiment was carried out using particle accelerators at the Ångström Laboratory at Uppsala University. Following tasks were completed before the experiment took place: (a) automation of the target position regulator, (b) development of control software, and (c) calibration and testing of the system. The deuterium dose was estimated at the level of1.9·1017 atoms/cm2.The deuterium concentration in molybdenum was found to be around 28·1015 atoms/cm2. This corresponds to a retention rate of around (15±3)%. / Fusion är en potentiell kandidat för hållbar kontinuerlig energi. Tyvärr är en fullt fungerande fusionsreaktor inte tillgänglig ännu och flera utmaningar kvarstår att lösa innan det blir en tillförlitlig källa. En av dessa utmaningar är plasma- inducerad modifikation av den inre väggen, dvs. strukturen närmast det heta plasmat i en tokamak. Tritium är en viktig komponent i ett framtida fusionsbränsle och väldigt sällsynt. Därför måste mängden bränsle som fastnar i väggen minimeras. I detta arbete har metoder för jonbestrålning av ett materialprov utvecklats. Med hjälp av det nya systemet har molybden (42Mo) bestrålats med deuterium (2H) och bibehållandet av deuterium har mätts med jonstråleanalys. Experimentet utfördes med hjälp av partikelacceleratorer i Ångströmlaboratoriet vid Uppsala Universitet. Följande uppgifter utfördes innan experimentet ägde rum: (a) automatisering av provmanipulatorn, (b) utveckling av programvara för styrning och (c) kalibrering och test av systemet. I ett avslutande test uppskattades den implanterade dosen till 1, 9 · 1017 atomer/cm2. Proverna var därefter analyserade och med kärnreaktionsanalys hittades ungefär 28 · 1015 atomer/cm2. Detta motsvarar ett bibehållnade på ungefär (12 ± 3)%. / Kandidatexjobb i elektroteknik 2020, KTH, Stockholm

<p dir="ltr">Research on enhancing the plasma confinement characteristics in fusion reactors and tokamaks has focused heavily on Low-Z plasma facing components Be, BeO, and SiC in recent decades. Building reactors, reducing harmful effects, and creating materials resistant to radiation all depend on an understanding of the plasma material interactions. In nuclear reactors, material composition and properties are also influenced by an understanding of impurity interactions. This thesis aims to investigate the effects of varying sputtering rates and long-term plasma durability on structured materials sputtered by plasma under various situations. The majority of this research has been done on the sputtering of materials as it accelerates the degradation of materials. To understand the process of ion solid contacts, a thorough investigation of ions' interactions with target atoms is presented in this work. Monte Carlo (MC) simulation has been done in this entire research by using the transport of ions in matter (TRIM). The influence of ion energy (100–1000 eV) and ion incidence angle by deuterium ions has been simulated in this study. As expected, on one hand, sputtering yield, as a function of ion-energy peaks first and a sequential reduction afterword; on the other hand, as a function of ion-incidence angle shows sequential enhancement towards max value followed by sharp reduction afterwards. The simulated data have been compared with the relevant experimental data and very close agreements were observed. To investigate the behavior of ion energy loss in relation to ion range in the targets, distribution profiles associated with ion range, recoil, ionization, and phonons are developed. Deuterium accumulation and its impact on Be target also have been shown in this work. The sputtering yield of <a href="" target="_blank">BeD₂</a> is lowest when the D incident ion interacts with low percentage of D has been simulated as target. Gradually increasing the percentage of D as target results in higher yields.</p>