Treat Reaktöründe Işınlanan Gelişmiş Metalik Alaşımların Radyasyon Hasarı Hesaplamaları
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The new generation reactor concepts that have been designed so far are foreseen for a common goal for safer, longer-lasting, nuclear non-proliferation and more economical use of nuclear energy. The most important issue to be considered in the process of successful development and application of new generation reactor systems is the problem of performance and reliability of structural materials used for applications inside and outside reactor vessel. Beyond the experience gained from existing nuclear power plants, structural materials need to be able to withstand much higher temperatures, higher neutron doses and high temperature corrosive coolant action. The materials considered to be active for use in different reactor components include various ferritic/martensitic steels, austenitic steels, nickel-based superalloys, ceramic composites and such materials. Operating conditions, such as high temperature, irradiation and corrosive environment, adversely affect material properties and create a risk of environmental damage, which can lead to high or severe consequences. Advanced metallic alloys including high-entropy alloys are considered to be promising structural materials for new generation reactors. In terms of superior radiation resistance, these alloys can also be selected as candidate materials for fusion technology. Innovations in the use of advanced metallic alloys include increasing the high temperature capacity and preventing irradiation-induced structural change. In recent years, with the help of these alloys, guidelines have been determined to overcome the problems that may arise in the use of these alloys. In this area, the elimination of data deficiencies and the development of new materials are important for the development of nuclear technology. Modeling material behavior is a promising tool to overcome long and expensive trial and error experiments. The modeling method, which will enable us to predict the behavior of materials, to be employed in this thesis constitutes one example of such method. The aim of this thesis is to determine the radiation damage parameters of nuclear alloys by modeling their behavior under radiation. For this purpose, neutronic characteristics of the TREAT reactor will be modeled using the SERPENT code. Upon checking the consistency of the obtained result with reference model results provided in the literature, radiation damage values for different advanced metallic alloys will be evaluated using SPECOMP and SPECTER codes.