Nükleer Enerji Mühendisliği Bölümü Tez Koleksiyonu
http://hdl.handle.net/11655/464
2020-10-28T09:36:53ZTreat Reaktöründe Işınlanan Gelişmiş Metalik Alaşımların Radyasyon Hasarı Hesaplamaları
http://hdl.handle.net/11655/22802
Treat Reaktöründe Işınlanan Gelişmiş Metalik Alaşımların Radyasyon Hasarı Hesaplamaları
Sarıoğlu, Burak
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.
2020-01-01T00:00:00ZDİNAMİK MONTE CARLO TEKNİĞİ İLE ZAMANA BAĞLI NÜKLEER REAKTÖR ANALİZİ
http://hdl.handle.net/11655/22751
DİNAMİK MONTE CARLO TEKNİĞİ İLE ZAMANA BAĞLI NÜKLEER REAKTÖR ANALİZİ
rashidian maleki, bahram; rashidian maleki, bahram
In this study, a detailed non-analog Dynamic Monte Carlo (DMC) methodology is provided and used to investigate the transient analysis of nuclear systems. The validity of the given methodology is demonstrated by solving a set of time-dependent benchmark problems. We also developed a new technique to generate time dependent Green's functions using DMC to perform transient analysis of source-driven systems (SDSs). The equilibrium and transient responses of SDSs are determined by using these generated Green's functions. This novel method is validated with comparison to different transient benchmark problem of SDSs.
In the conventional Point Reactor Kinetics Models (PRKMs), the time evolution of both neutron population and power are taken proportional to the weighted neutron density (amplitude function). The weight function must be chosen in such a way that the calculated kinetics parameters and amplitude function are as accurate as possible to characterize actual system. In this work, it is shown that, there are cases where the time dependency of power and neutron population differ. To overcome this problem, we derived the general forms of one- and two- point reactor kinetics models by using the actual neutron population and power, which are different from the conventional PRKMs. For different weight functions, the derived PRKMs are tested for transient analysis of one speed reflected slab reactors. Thus, obtained results are compared with the exact analytical solutions given by the Eigenfunction Expansion Method (EEM).
Furthermore, for different amount of reactivity insertions, taking into account the reactivity feedback, both non-analog dynamic Monte Carlo method named Point Kinetics Monte Carlo (PKMC) and Stochastic Point Reactor Kinetics Model (SPRKM) are developed, to simulate one- and two-point reactor kinetic models of the reflected reactors. Finally, a non-analog stochastic kinetics model is developed to simulate the PRKMs and compared with analog stochastic point kinetics model.
2020-09-02T00:00:00ZVVER-1200 Reaktöründe Soğutucu Kaybı Kazası ve Belirsizlik Analizi
http://hdl.handle.net/11655/22741
VVER-1200 Reaktöründe Soğutucu Kaybı Kazası ve Belirsizlik Analizi
Bilen, Osman
The objective of this thesis is to perform the thermal-hydraulic analysis of VVER-1200 reactor under Loss of Coolant Accident (LOCA) conditions including uncertainty calculations for the Peak Cladding Temperature (PCT) prediction using RELAP5/SCDAPSIM MOD 3.5 (RS MOD 3.5) best-estimate computer code.
In this work the emphasis has been given to the analysis of the performance of the Emergency Core Cooling System (ECCS) under large to intermediate break conditions. Two specific break sizes were studied: the Double-Ended Guillotine Break (DEBG), and the Transition Break Size (TBS). The selection of these break sizes emerged from the current 10.CFR.50.46 acceptance criteria, where DEGB is considered as a design basis accident (DBA), and particularly from the “proposed amendments to 10.CFR.50.46 providing risk-informed alternative LOCA break size”.
Within the scope of this thesis, an RS input model was generated based on the “Moscow NPP” design of VVER-1200 type reactor. The model was then used to observe the system behavior in the specified LOCA conditions namely for a 200% (double-ended, guillotine) cold leg break and for a 40.7% (single-ended, TBS) cold leg break. Furthermore, to account for the uncertainty in the PCT predictions, uncertainty calculations were carried out utilizing the integrated uncertainty package available in the RS MOD 3.5 code version.
The simulations performed in this study show that the ECCS performance is satisfactory in both accident scenarios. Therefore, we conclude that both hypothetical accidents can be tolerated in the VVER-1200 reactor. In the DEGB scenario the PCT reached a maximum of 948.1 K degree. In the TBS scenario, the system pressure loss rate is inferior to that in the DEGB case, resulting in a delay in the ECCS initiation. Nevertheless, smaller break results in a decreased mass of coolant loss rate and as a consequence decay heat removal were accomplished in a successful manner. The calculated PCT value never exceeded 627.7 K which is the cladding temperature during normal operation. Upon performing the uncertainty calculations, the upper limit for PCT was determined to be 1006.0 K under the DEGB scenario and 630.1 K for TBS. It is also concluded that the oxidation of the cladding is negligible and no remarkable hydrogen generation will result in both accidents.
2020-01-01T00:00:00ZNÜKLEER TIPTA KULLANILAN RADYOİZOTOPLARIN RADYASYON ZIRHLAMASI İÇİN YAZILIM GELİŞTİRİLMESİ
http://hdl.handle.net/11655/22678
NÜKLEER TIPTA KULLANILAN RADYOİZOTOPLARIN RADYASYON ZIRHLAMASI İÇİN YAZILIM GELİŞTİRİLMESİ
YURTTAŞ, NEŞE
In this study, a shielding software was developed in order to use in radiation shielding calculations of radioisotopes used in the Nuclear Medicine Departments of Hospitals in our country. Utilizing the software, it is aimed to easily calculate the thickness of concrete or lead shielding materials planned to be used in the areas where shielding is planned in Nuclear Medicine Departments.
C Sharp (C #) programing language is used for software. The results obtained from the developed software were compared with the calculations made with mathematical formulas and their correctness was checked.
Sample calculations were performed to determine the parameters need to be used in the software for different radionuclides and shielding materials by using the Monte Carlo method.
2020-07-07T00:00:00Z