The aim of the project is to design a steam generator in which the sodium flows pipeline hand, and on the other hand, water flow and water vapor (H2O). The main idea of this thesis is to propose such a steam generator, to avoid or eliminate the possibility of adverse reactions of sodium with water and, if this phenomenon occurs, thus minimizing the damage caused by this reaction. The work contains information on the implementation of steam, steam option, the necessary calculations for the design and construction and evaluation.

The goal of this thesis is to simulate the course of events during Dukovany Nuclear Power Plant blackout and to determine the optimal process of cooling down the cold branches of the primary circuit loops to secure the maximum amount of the primary fuel needed for the residual heat outlet so that the operating staff has as much time as possible for renewing the electric power supply. The first part of the thesis describes nuclear power plants built in the Czech Republic and in the world as well as reactor blocks whose construction is being considered. The detailed description of Dukovany’s steam generator is shared in chapter three. Following chapter summarizes blackouts that occurred at power plants around the world, events that might have led to blackouts in the Czech Republic, and it also describes current approach to blackout problematics at Dukovany Nuclear Power Plant. Chapters six and seven contain the core of the thesis. That includes detailed description of a mathematical model explaining the behaviour of a reactor block during blackout as well as the analysis of the results found.

The main aim of this thesis is the model design and control the secondary circuit block VVER 440. The search part of my work is a description of the secondary circuit. In the first part of the calculation is performed for the rated secondary circuit. The second calculation part focuses on the calculation circuit of the secondary for the reduced, influenced by shutting down one steam generator.

The aim of this work is to analyze the methodology for executing examination heat transfer tubes of steam generator by method of eddy currents on EDU, to create the model for the estimate of standard combined uncertainty of efficiency value, quantify the individual components of standard uncertainties, analyse the influence of different components on the expressing of uncertainty, to make the final estimate of standard combined and enlarged uncertainty of efficiency value and as of minor objective to evaluate the suitability of the use of this method .

In this thesis I review the sufficient flow capacity of the safety valve against high-pressure heater thermal pressurization in high-pressure feedwater heater on nuclear power Dukovany. I present basic possible cases of thermal pressurization both a high node heater, and the extended pressure unit in case of closure all power heads to steam generators.

The aim of this thesis is to address issues concerning the sealing of the lid for the primary collector within the steam generator (SG) of the nuclear power plant VVER 440. These steam generators are sealed in the original design by nickel rings. Modifying the existing method of sealing by a new type of sealing material, primarily from expanded graphite, can significantly reduce the pressure in the sealing surface and also stress in bolts and flanged joints. The new solution of sealing between the joint of collector and the lid should extend the life of joints and thus the nuclear and technical safety. The text is divided into the theoretical and computational part. A principle of the SG, the SG design, and a description of the joint and lid are mentioned in the theoretical part. The computational part shows calculations of the new joint, the original one, and comparison of both solutions in terms of technical and nuclear safety. The work contains drawing of the new joint.

The aim of the dissertation is proposing and solving energy convection systems for nuc-lear power plants with a sodium fast reactor of the 4th generation. The first part of the dissertation deals with collection and evaluation of information available about nuclear power plants with sodium fast reactor which use nuclear or non-nuclear reheating to increase thermal efficiency. On the basis of the acquired information, thermal schemes are developed and thermal effi-ciency is determined for the systems working in both Rankine thermal cycle and Brayton thermal cycle. In the further part of the dissertation thermal calculation of the reheater for nuclear and non-nuclear reheating is made for the systems working in Rankine thermal cycle. At the end of this dissertation, an apparatus suitable for these systems is suggested and the systems are evaluated in terms of technical implementation and nuclear safety.