Assessment of H trapping in W and W-based alloys: an atomistic study

• Location: Ugent (Belgium)
  
• Promotor: Prof. Guido Van Oost

• Situation: The technological feasibility of fusion reactors (DEMO) depends largely on the development of structural materials capable of withstanding prolonged irradiation with highly energetic neutrons, without seriously compromising their mechanical stability. Radiation effects in materials start at the moment a neutron reacts with the nucleus of an atom of the system and end with the degradation at the macroscopic scale of a structural component. Therefore, they are an inherently multiscale problem: phenomena occurring at very small scales, within very short timeframes, affect over the years the properties of tons of material. Computer simulations are nowadays a widespread and powerful tool to develop models describing the processes responsible for this degradation, at all scales. Our focus are Fe-Cr steels as structural materials and W-based alloys as armor and divertor material for fusion reactors. The proposed master thesis contributes to the effort of developing multiscale computer models for the description of the behaviour of the above specified materials under neutron irradiation. The work also falls in the framework of the European Fusion Development Agreement (EFDA). 
• Aims: Extended lattice defects such as dislocations and grain boundaries act as traps for plasma components penetrating the first wall. Hence, these defects serve as sources for nucleation of gas bubbles. As a result, He/H bubbles would grow faster in the grain boundary region causing its decohesion. Once the critical concentration of He/H at grain boundary is reached the material may fail by intergranular fracture. Hence, one needs to characterize the thermal stability of He/H bubbles at grain boundaries and their contribution to the decohesion. The objective of the proposed work is to use atomistic calculations (mostly molecular dynamics) to characterize the interaction of He and H with extended lattice defects such as dislocations and grain boundaries in W and Fe. The obtained results will be used to parameterize upper scale models (such as mean field theory) to understand the influence, implantation temperature, flux, dislocation density and grain boundary structure/size on the trapping of plasma components and corresponding formation and growth of He/H bubbles.

• Keywords:Radiation, fast neutrons, simulations, multiscale models, He/H bubbles
• Comments:
  • Supervisor: dr. Dmitry Terentyev [SCK.CEN] (Terentyev Dmitry (dterenty@sckcen.be) 
  • Location: Studie Centrum voor Kernenergie (SCK), Boeretang 200, B-2400 Mol 

Automated dust tracking in tokamak plasmas through analysis of fast camera images

• Location: Ugent (Belgium)
• Promotor: Prof. Guido Van Oost

• Situation: The formation and transport of dust particles is a common problem in fusion machines. The dust particles consist of agglomerates of smaller particles released from wall components through interaction with the plasma. If the dust propagates into the core plasma, it not only contaminates the plasma but it may also pose a serious safety problem for ITER and possibly cause severe performance issues for fusion reactors. Therefore it is essential to obtain information from experiments about the migration of dust particles through the plasma. Recently, fast cameras have been providing useful imagery of dust particle tracks in real time. From such images it is possible to determine statistics of the quantity and dynamics of dust particles, which is essential for a better understanding of dust formation and transport. 
• Aims: The aim of this thesis is to improve and implement a recent method for automatically tracking dust particles in video imagery, which was developed at the department. Improving the quality of the images will be addressed, as well as increasing the performance of the algorithm itself in identifying individual dust particles and tracking them over a maximum portion of their trajectory. Measurements will be available from the ASDEX Upgrade tokamak (IPP, Garching) and from the linear plasma device Mirabelle (UHP, Nancy) For the interested student, a visit to the TEXTOR tokamak (FZ-Juelich, Germany) can be arranged.

• Keywords:dust transport, ITER, automatic tracking, Asdex upgrade,

• Comments:
  • Supervisor:Geert Verdoolaege (Geert.Verdoolaege@UGent.be) 
  • Location:: Ugent

Development of new measurement methods and techniques for the ITER fusion research project

• Location: Ugent (Belgium)
• Promotor: Prof. Guido Van Oost

• Situation: Fusion energy powers the sun and the stars of our universe. The aim of nuclear fusion research is to master this energy source on earth and, ultimately, to provide for the benefit of all, an almost limitless source of energy with highly attractive safety and environmental features. The international ITER fusion research reactor, currently under construction in Cadarache, South of France, will require sophisticated and dedicated instrumentation and measurement techniques capable to operate in the extreme conditions created by the hot fusion plasma core (>100 million kelvin). In ITER, magnetically confined fusion alpha particles which are the products of the Deuterium –Tritium fusion reactions and which carry about 20% of the fusion power play a key role of sustaining the plasma core temperature. Developing measurement techniques for fusion alpha particles is, therefore, crucial for the ITER project.
• Aims: This project is a contribution to the design and development of new instrumentation for the metrology of fusion alpha particles. New detectors based on nuclear reactions are designed and their response calculated. Modelling is done with existing codes and/or developing new tools. The obtained results may also be applied to the interpretation of data from existing fusion experiments such as JET and ASDEX[1]. [1] JET ,the Joint European Torus is the current world’s largest fusion experiment, located near Oxford, United Kingdom. ASDEX is Germany’s largest fusion experiment at the Max-Planck-Institute located near Munich

• Keywords:ITER, fusion,magnetic confinement, deuterium, plasma

• Comments:
  • Supervisor: dr. George Bonheure 
  • Location: ERM/KMS Brussels

Electrostatic probes for the study of plasma parameters and turbulence in the boundary of tokamaks

• Location: Ugent (Belgium)
• Promotor: Prof. Guido Van Oost

• Situation Tokamaks are the most advanced devices in magnetic fusion research. The first experimental reactor ITER will be a tokamak. Electric fields and their shear play a very important role in the suppression of turbulent losses ( the main loss mechanism in tokamaks) and in the establishment of improved confinement schemes ( "advanced tokamak scenarios). For the study of plasma parameters and electrostatic turbulence in the boundary of the discharge electrostatic probes can be used. For that purpose advanced electrostatic probes have been developed and tested on the small tokamak CASTOR in Prague. The success lead to application in larger devices. Aims The aim is to contribute to the experimental programme and data analysis of probe (different types) measurements in the tokamak COMPASS (scale model of ITER) of the Institute of Plasma of the Academy of Sciences in Prague.

• Keywords:data analysis, COMPASS, ITER, plasma, tokamak

• Comments:
  Location: Prague, Technicum 
  www.cas.cz A preceding internship during the Summer will be necessary. This can be combined within an Erasmus exchange programme. Accommodation will be organized locally.

Neutral beam heating on the tokamak COMPASS

• Location: Ugent (Belgium)
• Promotor:Prof.Guido Van Oost

• Situation The COMPASS tokamak, originally from CCFE (former UKAEA) Culham in United Kingdom, was re-installed in the institute of plasma physics ASCR, in Prague, Czech Republic. The tokamak vacuum vessel is a D-shaped one, with the main radius R = 0.56 m and the minor radius a = 0.2 m. Toroidal field BT is in the range of 0.8 – 2.1 T. Interesting feature of the tokamak COMPASS is its ITER-geometry, in 1:10 scale. This offer an opportunity to add the COMPASS into the row of tokamaks with progressive size: COMPASS, ASDEX-U, JET, ITER. This row of geometrically relevant tokamaks is important for scaling of experimental results from the first three of them to ITER . Another interesting feature COMPASS is the possibility to reach the high confinement mode or so called H-mode (in comparison to the “normal” Low-mode) thanks to the presence of the two NBIs. COMPASS could theoretically reach the H mode for hydrogen alone at the maximum power, but the presence of NBI is necessary to reach it with He or D. 
• Aims Study of plasma heating in tokamak is of great importance in the field of magnetically confined fusion. The two Neutral Beam Injectors designed for the COMPASS tokamak will contribute to the worldwide effort to deepen the knowledge of high energy heating source and plasma-beam interaction, for the purpose of fusion plasma heating. The main aim of the thesis is the study of the interaction between the neutral beam and the plasma of the tokamak COMPASS.

• Keywords:COMPASS, Tokamak

• Comments:
  
  • Location: Institute of Plasma Physics Prague & Technicum A preceding internship during the Summer will be necessary. This can be combined within an Erasmus exchange programme. Accommodation will be organized locally.

Real-time identification of ELM types and disruptions in fusion plasmas through classification in information spaces

• Location: Ugent (Belgium)
• Promotor: Prof. Guido Van Oost

• Situation In present-day tokamak plasmas it is essential to study instabilities such as edge-localized modes (ELMs) and disruptions. Indeed, these phenomena will have to be controlled or, in the case of disruptions, eliminated in next-step devices such as ITER and beyond. Based on their causes and characteristics, ELMs and disruptions can be classified into various types and it is important to be able to identify the types. Indeed, provided the identification and prediction of events is possible in real time, this contributes valuable information for the operation and control of the device. One approach for characterizing ELMs and disruptions is data-driven, i.e. a learning machine is trained to recognize the patterns that define an instability and to classify the event accordingly. Due to the stochastic nature of the plasma events and diagnostic signals, a probabilistic approach is required. Furthermore, recent experiments concerning the identification of plasma confinement modes and prediction of disruptions suggest that the probability distribution corresponding to a characterizing physical quantity carries significantly more useful information, compared to the measured value itself. This information must be exploited for an accurate and robust prediction and classification of instabilities. The challenge, however, is to develop classification techniques that are able to deal directly with probability distributions. This can be based on a natural similarity measure between distributions, given by the geodesic distance in the corresponding 
• Aims: In this thesis, classification algorithms will be developed that are able to discriminate between probability distributions and that are hence well suited for discovering and learning patterns in stochastic plasma data. Measurements from the JET tokamak (CCFE, Oxford) will be available. Depending on the interests of the student, focus may be put on the computational aspects or rather on the implementation of algorithms, with a view to real-time analysis at JET. For the interested student, a visit to the TEXTOR tokamak (FZ-Juelich, Germany) can be arranged.

• Keywords:TEXTOR, Tokamak, disruptions, ELM, stochastic plasma,
• Comments:
  • Supervisor: Geert Verdoolaege (geert.verdoolaege@ugent.be)

Visualization of tokamak operational spaces through the projection of data probability distributions 

• Location: Ugent (Belgium)
• Promotor: Prof. Guido Van Oost

• Situation The operational space of the plasma in a fusion device is essentially a multi-dimensional state space wherein the plasma state is represented by the measurement of a number of key physical and engineering quantities. Exploration of tokamak operational spaces is an essential activity in fusion research for establishing the conditions under which specific operational regimes and plasma instabilities develop. An important way to realize this level of understanding is through visualization of the plasma state in a 2D map. The main value of such a map is as a tool for visualization of data and key regions of the operational space. On the one hand, this provides physicists with a convenient means to study data patterns (relationships, clusters) reflecting specific regimes and their underlying physics. On the other hand, the plasma state at any given time can be mapped into such a diagram. This conveys important information regarding the proximity of the plasma to operational regimes that are either desired or to be avoided. In addition, the trajectory of the plasma, as it moves through part of the operational space during a discharge, can be visualized as well. 
• Aims: In this thesis a recent technique for projecting high-dimensional plasma data into a 2D map will be implemented as a demonstration tool for use in the control room of fusion machines. The projection technique will be based on the well-known multidimensional scaling algorithm. However, since the measurements of plasma quantities are typically highly stochastic, they will be modeled with probability distributions. The projection technique will thus employ a distance measure between distributions, which in this case will be the Rao geodesic distance. Data will be available from the ITPA Global H-Mode Confinement Database and from the JET tokamak. The software tool should be able to visualize the instantaneous plasma state in a diagram, indicating the primary operational regimes and dangerous (instable) regions. The software will have to be capable of (near) real-time operation. Another aim is to map the trajectory of the plasma during a discharge and this will be done for several JET plasma pulses. For the interested student, a visit to the TEXTOR tokamak (FZ-Juelich, Germany) can be arranged.

• Keywords:TEXTOR, Tokamak, high-dimensional plasma data,
  
• Comments:
  • Supervisor: Geert Verdoolaege (geert.verdoolaege@ugent.be)  

Experimental study of nitrogen scavenging to reduce hydrocarbon deposition on fusion reactor walls (T3)

• Location: Ghent University (Belgium)
• Promotor: Christophe.Leys@UGent.be
• Tungsten tiles with carbon fibre composites are among the candidate materials for the divertor of ITER. The use of carbon however can be problematic due to the redeposition of tritium-containing carbon layers in hidden regions of the divertor. One approach to reduce this redeposition is to inject nitrogen as a scavenger molecule that transforms reactive species into non-reactive ones. The technique also has an advantage because of the plasma cooling effect due to energy exchange to N2. The use of N2 results in to suppression of carbon films growing but formation of tungsten nitrides can strongly effects the operation of the system. Description: A laboratory experiment will be set up to explore the scavenging mechanism and formation of WN3 in a low temperature plasma containing H2, CH4 and nitrogen. The RF-excited low pressure plasma will be investigated by optical emission spectroscopy of ions and neutral species and mass-spectroscopy. The formation of carbon and nitride layers on substrates mounted on the electrodes with heater will be investigated for different working conditions (gas mixture, power, substrate temperature). The goal is to establish a relation between the plasma parameters and the formation of films on the surface of W.
• Keywords:Keywords:plasma, fusion, ITER, experiment, spectroscopy

How do grain boundaries affect the strength of tungsten alloys for fusion reactors ?

• Location: Ugent (Belgium)
• Promotor: Stefaan Cottenier
  
  
• Most design proposals for fusion reactors contain tungsten or tungsten-based alloys. The reason for this is the high melting temperature of tungsten, combined with its low activation. A disadvantage of tungsten is its limited ductility, which makes it problematic to use tungsten for structural applications. Materials scientists therefore wonder whether it is possible to develop tungsten alloys that conserve the high melting temperature, yet provide a better ductility. Several reports on experimental studies on W-alloys are available in the literature. None of those alloys turns out to be particularly ductile, so far. The number of candidate alloys that has not been examined yet, is large. Experimental studies are time-consuming, and therefore progress will be slow. One way to speed this up, is to use atomistic modeling instead. A tungsten alloy can be subject to a virtual atomistic tensile test, which can reveal information that is relevant to the macroscopic tensile properties of the alloy. In this master thesis, you will use ab initio codes to perform tensile test simulations on tungsten samples. You will study how grain boundaries affect ductility and tensile strength. This will be done for pure tungsten, as well as for tungsten containing alloying elements at the grain boundary. The results might guide future experimental research. Additional info: Rieth et al., J. Nucl. Mater. 432 (2013), 482-500, http://dx.doi.org/10.1016/j.jnucmat.2012.08.018 Lazar en Podloucky, Phys. Rev. B 78 (2008), 104114, http://dx.doi.org/10.1103/PhysRevB.78.104114
  
  
• Keywords:computational materials physics, grain boundaries, simulated tensile testing

High-throughput computation: vanadium alloys for DEMO.

• Location: Ugent (Belgium)
  
• Promotor: Stefaan Cottenier
  
  
• Steel is ubiquitously used as a structural material in industrial constructions. For application in the specific environment of a fusion plant, however, low activation structural materials are required. In common structural materials, the neutron flux of the reactor would lead by transmutation to a large amount of long-lived isotopes. The main candidate for low activation structural applications in DEMO are vanadium-based alloys (see Refs. [1] and [2] for reviews). The alloy that has been studied most, is V-4Cr-4Ti (bcc vanadium with 4% chromium and 4% titanium). In order to come to this conclusion, many vanadium alloys have been tested experimentally. It is too time-consuming to examine all possible alloys by experiments only. It becomes increasingly more feasible, however, to tackle such a task by means of ab initio computer simulations. This is called high-throughput computation [3]: calculating properties of many candidates, putting these into a data base, and examining the results with data mining tools. In this master thesis, you will examine the model system V-6X-6Y (bcc vanadium with 6% of element X and 6% of element Y), where X and Y are any of the ten 3d transition metals. Using an ab initio code, you will calculate properties of a few hundred representative alloys. You will construct a data base of these, and analyze it with data mining methods. This can result in specific suggestions to experimental scientists about which alloys are most promising to study in the lab. [1] Votinov et al., Journal of Nuclear Materials 233-237 (1996) 370 [2] Tavassoli, Journal of Nuclear Materials 302 (2002) 73 [3] The high-throughput way to computational materials design, S. Curtarolo et al., http://www.nature.com/doifinder/10.1038/nmat3568
  
  
• Keywords:computational materials design,high-throughput computation,vanadium alloys

Studies of protection techniques for ITER diagnostic mirrors

• Location: Ugent (Belgium)
• Promotor: Prof.Guido Van Oost
• Metallic mirrors will be used in all optical and laser diagnostic systems in ITER, viewing plasma radiation in the wavelength range from a few nm to a hundred of microns. In ITER, mirrors may become contaminated with impurities from burning plasma. Such a contamination will degrade the reflectivity of mirrors, thus negatively affecting the entire performance of respective diagnostic systems. Therefore, the development of efficient techniques for mirror protection from contaminants is of prime importance for ITER optical diagnostics. The most of first mirrors will be installed In ITER at the end of diagnostic ducts at various distances from the plasma. The predictive modeling and first experiments show that protection of the entrance of diagnostic duct with a shutter may decrease adverse impurity fluxes towards the mirrors. In addition, special shaping of the diagnostic duct may efficiently trap impurities moving towards the mirrors, thus increasing the mirror lifetime. To validate the efficiency of shutters and duct geometry, several mirror systems equipped with diagnostic tubes and single-crystal molybdenum mirrors at their end, were exposed in several tokamaks: DIII-D (USA), ASDEX Upgrade (Germany, EU) and TEXTOR (Germany, EU). Some of diagnostic tubes were protected with shutters, some tubes had inner apertures to trap impurities on their way to the mirrors. The present work will be devoted to evaluation of the efficiency of mirror protection in the exposed mirror systems. The work will comprise surface analyses and optical measurements on mirrors and in the diagnostic ducts. Measurements of total and diffuse reflectivity will be made in the state-of-the-art MirrorLab in Jülich. Electron- and ion-beam surface analyses available at Forschungszentrum Jülich (FZJ) and in the partner laboratories will be used to analyze the amount of impurities. At the end of master course, the efficiency of the applied mirror protection techniques should be assessed. The work will be made in collaboration with involved tokamak teams.
• Keywords:mirrors, ITER, contamination, molybdenum
• Comments:
  • Supervisor: Dr. A. Litnovsky,
  • Localisation: FZ Jülich