| 1. |
- Garcia Carrasco, Alvaro, 1989-
(författare)
-
Impact of material migration on plasma-facing components in tokamaks
- 2016
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- 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.
|
|
| 2. |
- Ivanova, Darya, 1983-
(författare)
-
Plasma-Facing Components in Tokamaks : Material Modification and Fuel Retention
- 2012
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- 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.
|
|
| 3. |
- Moon, Sunwoo, 1984-
(författare)
-
Impact of erosion and deposition processes on wall materials in tokamaks
- 2022
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Understanding of material migration and control of fuel retention are essential for the safe operation of a reactor-class fusion machine. Work presented in the thesis focuses on erosion-deposition processes which are decisive for the formation and properties of co-deposited fuel-containing layers on plasma-facing and diagnostic components, and for the dust formation. The thesis is based on experiments carried out in plasma devices such as JET-ILW, KSTAR, EXTRAP-T2R, TOMAS and, in materials research laboratories where comprehensive analyses of the plasma-exposed materials were performed by a large number of complementary ion, electron and optical methods. The major objectives were to determine: (a) plasma impact on test mirrors; (b) properties of metal dust generated under operation with metal walls in JET with the ITER-Like Wall; (c) material transport to areas shadowed from the direct plasma line-of-sight; (d) neutral particle fluxes in wall conditioning discharges. All these topics are inter-related and, they are in line with the ITER needs in areas of diagnostic development, mitigation of fuel inventory and detailed knowledge of dust particles generated in the tokamak with metal walls. The novelty in research is demonstrated by several elements. Plasma impact on diagnostic mirrors was determined by exposure of test mirrors in JET two types of holders including the ITER-like assembly resembling a diagnostic duct in a reactor. Dust studies allowed for the determination of particles’ properties (size, weight) and, for the classification of various detected objects. The impact of tile shaping and intentional misalignment on fuel retention was revealed in a dedicated experiment in KSTAR. A neutral particle analyser was first tested at EXTRAP-T2R and then installed at the TOMAS facility. Particle fluxes were characterized in wall conditioning discharges heated by electron- and ion cyclotron systems.
|
|
| 4. |
- Petersson, Per, 1978-
(författare)
-
Ion Beam Analysis of First Wall Materials Exposed to Plasma in Fusion Devices
- 2010
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- One major step needed for fusion to become a reliable energy source is the development of materials for the extreme conditions (high temperature, radioactivity and erosion) caused by hot plasmas. The main goal of the present study is to use and optimise ion beam methods (lateral resolution and sensitivity) to characterise the distribution of hydrogen isotopes that act as fuel. Materials from the test reactors JET (Joint European Torus), TEXTOR (Tokamak Experiment for Technology Oriented Research) and Tore Supra have been investigated. Deuterium, beryllium and carbon were measured by elastic recoil detection analysis (ERDA) and nuclear reaction analysis (NRA). To ensure high 3D spatial resolution a nuclear microbeam (spot size <10 µm) was used with 3He and 28Si beams. The release of hydrogen caused by the primary ion beam was monitored and accounted for. Large variations in surface (top 10 µm) deuterium concentrations in carbon fibre composites (CFC) from Tore Supra and TEXTOR was found, pointing out the importance of small pits and local fibre structure in understanding fuel retention. At deeper depths into the CFC limiter tiles from Tore Supra, deuterium rich bands were observed confirming the correlation between the internal material structure and fuel storage in the bulk. Sample cross sections from thick deposits on the JET divertor showed elemental distributions that were dominantly laminar although more complex structures also were observed. Depth profiles of this kind elucidate the plasma-wall interaction and material erosion/deposition processes in the reactor vessel. The information gained in this thesis will improve the knowledge of first wall material for the next generation fusion reactors, concerning the fuel retention and the lifetime of the plasma facing materials which is important for safety as well as economical reasons.
|
|
| 5. |
- Ström, Petter, 1989-
(författare)
-
Material characterization for magnetically confined fusion : Surface analysis and method development
- 2019
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- The dream of abundant clean energy has brought scientists and laypeople alike to ponder the possibilities of nuclear fusion since it was established as the energy source of the stars in 1939. Starting from the mid-twentieth century, significant effort has been put into overcoming the technological challenges related to the construction of a power plant, but initial optimism has faded somewhat due to a notable absence of practical outcomes. Nevertheless, the research continues and progress is made slowly but surely.The present work deals with a small part of the fusion puzzle, namely the materials to be used in the first wall surrounding a magnetically confined plasma. Carbon, which has historically been considered as the most viable element for this role, has been ruled out due to issues with plasma-induced erosion, hydrocarbon formation and a buildup of thick deposited material layers on wall components. The latter two lead to an unacceptable accumulation of radioactive tritium, both in the deposited layers and in dust particles. A metal wall, which would alleviate these particular problems but increase the severity of others, is therefore envisioned for a future demonstration reactor.Three contributions to the overall research effort are made through this thesis. First, an increased understanding of plasma-induced erosion of so-called reduced activation ferritic-martensitic steels and preferential sputtering of light material components is provided. High-resolution ion beam analysis and microscopy methods are used to examine samples of such a steel after exposure to plasma under controlled circumstances. Model films consisting of a mixture of iron and tungsten deposited on silicon substrates are also studied as they constitute simpler systems where the effects of interest may be simulated. The knowledge obtained is necessary for an assessment of the possibility to use reduced activation steel as a plasma-facing material in specific regions of a reactor wall.The second contribution consists of reports on the composition of deposited material layers on wall components retrieved from the plasma confinement experiments JET and TEXTOR. These provide limited conclusions on the range and rate of material erosion, transport and deposition in two cases.Finally, a detection system for the ion beam technique elastic recoil detection analysis has been assembled, tested and put into operation. In addition to improving the quality of analyses performed on fusion-related materials, the system has become an established tool available for users of the 5 MV electrostatic pelletron accelerator at Uppsala University’s Tandem Laboratory.”
|
|
| 6. |
- Weckmann, Armin
(författare)
-
Material migration in tokamaks : Erosion-deposition patterns and transport processes
- 2017
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Controlled thermonuclear fusion may become an attractive future electrical power source. The most promising of all fusion machine concepts is called a tokamak. The fuel, a plasma made of deuterium and tritium, must be confined to enable the fusion process. It is also necessary to protect the wall of tokamaks from erosion by the hot plasma. To increase wall lifetime, the high-Z metal tungsten is foreseen as wall material in future fusion devices due to its very high melting point. This thesis focuses on the following consequences of plasma impact on a high-Z wall: (i) erosion, transport and deposition of high-Z wall materials; (ii) fuel retention in tokamak walls; (iii) long term effects of plasma impact on structural machine parts; (iv) dust production in tokamaks.An extensive study of wall components has been conducted with ion beam analysis after the final shutdown of the TEXTOR tokamak. This unique possibility offered by the shutdown combined with a tracer experiment led to the largest study of high-Z metal migration and fuel retention ever conducted. The most important results are: - transport is greatly affected by drifts and flows in the plasma edge;- stepwise transport along wall surfaces takes place mainly in the toroidal direction;- fuel retention is highest on slightly retracted wall elements;- fuel retention is highly inhomogeneous. A broad study on structural parts of a tokamak has been conducted on the TEXTOR liner. The plasma impact does neither degrade mechanical properties nor lead to fuel diffusion into the bulk after 26 years of duty time. Peeling deposition layers on the liner retain fuel in the order of 1g and represent a dust source. Only small amounts of dust are found in TEXTOR with overall low deuterium content. Security risks in future fusion devices due to dust explosions or fuel retention in dust are hence of lesser concern.
|
|
