Thermonuclear fusion offers the possibility of a safe, abundant, clean and sustainable source of energy using small amounts of fuel (hydrogen isotopes). High temperatures as well as high fluxes of neutrons and other highly energetic particles during the operation of a magnetic confinement plasma device (Tokamak) make imperative the use of materials that are resistant to high temperatures and radiation. Tungsten (W) is a candidate plasma facing material for the interior walls of a fusion reactor due to its high melting point, high thermal conductivity, low tritium retention and heat stress resistance. However, tungsten suffers from brittle behavior at low temperatures due to the relatively high ductile to brittle transition temperature (DBTT) which ranges between 400 and 700 K, limiting its exploitation.

The objective of Vasileios Chatzikos’ PhD is to study the transmutation products of tungsten as well as the created open volume defects as a function of the irradiation temperature, the neutron irradiation dose and the tungsten microstructure. The experimental methods used to achieve this investigation are gamma-spectroscopy and positron annihilation lifetime spectroscopy (PALS).
Three types of tungsten are under investigation a) single crystal tungsten W (100), b) cold rolled tungsten sheet and c) double forged tungsten in the form of a square bar (ITER grade). These materials have been irradiated at the BR2 research reactor in Belgium, at four temperatures (600, 800, 900 and 1200 C) and at irradiation doses corresponding to 0.1, 0.18, 0.5 and 0.75 displacements per atom (dpa).

The experimental results of gamma-spectroscopy measurements on tungsten samples lead to the identification and determination of the radioactive isotopes of Tungsten, Rhenium and Tantalum concentrations as a function of the neutron irradiation dose. Through the comparison of the experimental values with theoretical calculations, using MCNP neutron transport code and FISPACT-II radionuclide inventory code, the quantification of the transmutation products in the irradiated tungsten is achieved.
Employing Positron Annihilation Lifetime Spectroscopy (PALS) spectroscopy technique the evolution of the open volume defects type (vacancies, dislocations and vacancy clusters) and their relative concentrations is determined as a function of irradiation temperature and dose. Moreover, the evolution of the open volume defects will be assessed in correlation with the transmutation products.

Understanding the radiation effects of energetic ions and neutrons on steel has considerable scientific interest and additionally important technological impact related to energy production by fusion. These two aspects are interrelated, as a basic understanding of the phenomena will promote the development of radiation resistant materials, necessary for the implementation of the international plans for a fusion reactor in the next twenty years.

Ferritic–martensitic steels based on Fe–Cr alloys are candidate materials for the structural components of fusion power plants. These materials will be exposed to extremely high neutron fluxes and elevated temperatures.

Self – ion irradiation can simulate the neutron – induced radiation damage in materials, because the dominant damage arises from Primary Knock-on Atoms (PKA) in which the neutron energy is transferred.

Decomposition occurring in FeCr alloys during thermal aging is typically occurring between 773 and 813 K, temperatures at which the thermal diffusion is adequate to drive it. Under irradiation, short-range ordering or decomposition can be observed at even lower temperatures, because the point defect supersaturation accelerates diffusion processes i.e. radiation induced segregation (RIS). RIS is the process by which the composition of an alloy is altered due to preferential participation of certain species with the vacancy and/or interstitial flux to sinks.

Τhe objective of the PhD thesis is the study of the effect of iron ion irradiation on the properties of Fe-Cr films as a function of Cr content, irradiation dose and temperature, damage rate and iron ion energy. The focus will be on the determination of the global equilibrium conditions which are reflected by the solute Cr content in the matrix after irradiation. Also the observed phenomena will be related to the mechanical behavior of the FeCr films under irradiation.

High temperatures, as well as high fluxes of neutrons and other highly energetic particles during the operation of a magnetic confinement plasma device (Tokamak) make imperative the use of materials that are resistant to high temperatures and radiation. Tungsten is a candidate plasma facing material for the interior walls of a fusion reactor due to its high melting point, high thermal conductivity, low tritium retention and heat stress resistance. However, tungsten suffers from brittle behavior at low temperatures due to the relatively high ductile to brittle transition temperature, limiting its exploitation.

The objective of Dimitris Papadakis’ PhD is the study of the annealing effects of fission neutron irradiated tungsten materials.

Selected neutron irradiated tungsten materials will be annealed in the temperature range of 600 – 1500 C with the aim of studying the restoration of the properties of tungsten to the pre-irradiated state. Neutron irradiation causes damage in a material, resulting in the creation of defects. With the use of Positron Annihilation Lifetime Spectroscopy the type and percentage of open volume defects will be determined. Additionally, optical and electron microscopy (SEM), as well as X-ray diffraction (XRD) will be used to determine the changes in the structure of the materials during annealing and irradiation.

Annealing at very high temperatures results in structural changes of the tungsten material through diffusion and the interaction of various defects in the crystalline lattice, or even through the mechanism of recrystallization. These structural changes lead into the change of the mechanical and electrical properties of the materials. Through various techniques of characterization of mechanical properties (Macro-Nano Indentation, Impulse Excitation) and electrical properties (DC Resistivity), structural changes will be correlated with changes in the physical properties of the material.

The PhD thesis will be defended at the Physics Department of UoA.

Objective of the project is the study of the interactions between radiation defects and solute atoms in ferritic steels based on the binary Fe‑Cr system. The experimental results will be compared with theoretical prediction in order to validate theoretical models.

Ferritic steels are considered the main choice as structural materials for the future fusion power plants. The impurities solute atoms such as carbon (C) and nitrogen (N) are main components in steels that define their mechanical and thermal properties and play also an important role to the configuration of their microstructure. During the ion irradiation the interaction of energetic particles with matter can cause atomic displacements leading to several types of defects such as vacancies and interstitial atoms. After their creation defects can diffuse and interact with solute impurities resulting to changes in microstructure and therefore to macroscopic properties.

In order to study the radiation defects, samples of ferritic alloys are irradiated by protons at NCSR “Demokritos” TANDEM accelerator at the facility IR2 [link: infrastructure/IR2]. The irradiations are performed at cryogenic temperature so the defects are immobile into the sample. Subsequently, through a post-irradiation thermal annealing process of the samples the defects can diffuse and their evolution is observed by in-situ electrical resistivity measurements. Information about the interactions between radiation defects and impurities can be revealed by comparing the resistivity recovery spectrum of samples with different impurity concentration.

Interaction between plasma and plasma facing materials is an issue of great importance for the safe operation of the fusion devices. Beryllium (Be) is a candidate material to be used in the main chamber of ITER and the future fusion devices based on magnetically confined plasma. Beryllium due to its low atomic number prevents the dilution of plasma and presents low fuel retention which is crucial for the life time of the wall and the conservation of the fuel. Moreover, beryllium is an oxygen getter which reduces oxygen impurities inside the vessel.

The objective of Pavlos Tsavalas’ PhD is the investigation of plasma facing material from the JET tokamak main chamber after the interaction with the plasma in order to assess material erosion, fuel retention and material deposition from other areas of the tokamak. In order to achieve this investigation, the following methods have been used:

1. Ιon beam analysis with deuteron (2H) and helium (3He) to detect, quantify and assess the depth profile of the light elements (deuterium, beryllium, carbon, nitrogen and oxygen) and the micro-beam to depict the mapping of the same elements on the surface.
2. Differential cross sections measurements of deuteron reaction on beryllium which are necessary for the quantitative results of the ion beam analysis.
3. X-Ray fluorescence to assess the relative concentration of the heavier elements (chromium, iron, nickel, molybdenum and tungsten) in the whole volume of the samples.
4. Scanning electron microscopy with energy dispersive spectroscopy of X-rays to investigate the mapping of the heavier elements and the morphology of the surface.
5. X-Ray diffraction to assess any compound formation in the samples.

The PhD thesis will be defended in at the School of Applied Mathematical and Physical Sciences of National Technical University of Athens.