National Repository of Grey Literature 41 records found  1 - 10nextend  jump to record: Search took 0.00 seconds. 
The design of components for emergency cooling of reactor pressure vessel
Katzer, Milan ; Šen, Hugo (referee) ; Martinec, Jiří (advisor)
My thesis deals with the design of an experimental emergency cooling device of the reactor pressure vessel (RPV). It consists of two parts, the theoretical one and practical one. Different molten corium cooling methods in terms of their efficiency and comparison are introduced in the theoretical part. The design of an experimental emergency cooling device, which incorporates a model channel past the reactor pressure vessel , is presented in the practical part. The cooling device consists of a model channel past the reactor pressure vessel, condensator, which takes away the heat generated by the reactor pressure vessel and the pump of a secondary loop. Next, thermal and hydraulic calculations are given in this section. The conclusion is devoted to the evaluation of particular cooling technologies and their comparison in terms of nuclear and technical safety.
The steam generator for ESFR reactor
Bátěk, David ; Martinec, Jiří (referee) ; Šen, Hugo (advisor)
This master thesis deals steam generator for ESFR (European Sodium Fast Reactor), which is heated by liquid sodium. In the beginning chapters, there are theoretic information about ESFR's parameters and its' comparison with ohter types of heat exchangers in nuclear reactors with the same principal (sodium as a coolant). Then designing part follows, which contents of introduction of calculations, option of material and conception of heater. Computational part on its own includes thermal, hydraulic and stress calculations and comparison with aspects in nuclear safety and security.
Design of the steam generator for modular reactor
Černý, Marian ; Baláš, Marek (referee) ; Šen, Hugo (advisor)
Subject of this thesis is design of the steam generator for modular reactor. The dissertation consist of the theoretical part, where are described heat-exchangers and steam generators used in nuclear power plants. Second part contains theoretical calculations in the first approach (thermal, hydraulic and strenght calculation). In the next part are particular variants of steam generator selected. For the final variant are necessary calculations (introduced above) and drawings of selected parts are done. In the final statement is technical solution evaluated, and the parameters of the steam generator are compared with actually used steam generators.
The flange gasket modification of SG VVER 440 primary collector
Blažková, Eva ; Martinec, Jiří (referee) ; Šen, Hugo (advisor)
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 flange gasket modification of SG VVER 1000 primary collector
Pransperger, Jan ; Martinec, Jiří (referee) ; Šen, Hugo (advisor)
This thesis deals with the replacement of the lid gasket of the primary collector of the steam generator VVER 1000. Original sealing by nickel rings is replaced by kammprofile gasket with expanded graphite layers. The thesis compare the properties of both types of gaskets and the new and the original configuration of flange joint which have been calculated according to EN 1591. The results are compared and conclusions arising therefrom are presented. The work includes results of FEM analysis of the new configuration of the flange joint. There is also a description of the main components of the nuclear power plants VVER 1000 primary circuit in the introductory part which focused on the construction and operation of the steam generator and its primary collector.
The Intermediate Heat Exchanger for ESFR reactor primary circuit
Švihel, Miroslav ; Baláš, Marek (referee) ; Šen, Hugo (advisor)
The thesis is mainly focused on the design of the intermediate heat exchanger primary circuit of the reactor ESFR. Heat exchanger is calculated heat, hydraulic and strength and is finally processed part drawings. There are designed the basic dimensions of the tube bundle and container heat exchanger. There are included an overview of concepts and so far used types IHX at the nuclear power plants with fast reactors. There are also mentioned basic parameters of the project ESFR and evaluated the safety and operational reliability of the heat exchanger.
Radiation damage of VVER-440 based Dukovany NPP reactor pressure vessel investigation
Říha, Tomáš ; Martinec, Jiří (referee) ; Šen, Hugo (advisor)
This master‘s thesis deals with radiation damage of reactor pressure vessels, specifically of NPP Dukovany Unit No. 3. In general damage mechanisms of reactor steels are described and possibilities of monitoring of material degradation and its recovery used at NPP’s all over the world are mentioned as well. A practical part of the thesis is focused on interpretation of analyses carried out with the assistance of MOBY DICK code. The ground of these analyses is a neutron fluence value development on different locations of RPV for the whole life of operation up to 24th cycle. The analyses results are put into context with performed in-service inspections. The thesis follows up with neutron fluence computation for the future cycles containing new types of nuclear fuel up to 34th cycle. The outcome of practical part of the master‘s thesis is a comparison between new types of nuclear fuel with respect to radiation damage of RPV’s.
Spent fuel storage casks thermal and physical properties investigation
Hlatký, Pavel ; Martinec, Jiří (referee) ; Šen, Hugo (advisor)
This work deals with questions of spent fuel storage casks thermal and physical properties investigation. Foundations of mathematics which are necessary for describing field of temperature are included. The work itself contains calculation methods which are split into two parts. The first one deals with simplified analytic solution and the second part solves the whole problem by the numerical computation.
The flange gasket modification of TC SVO1 ion filter manhole on the VVER 440 NPP
Šnajdárek, Ladislav ; Fiedler, Jan (referee) ; Šen, Hugo (advisor)
This diploma thesis is engaged in replacement of gasket ion filter used in the ion filter of continuous cleaning TC SV01 of rector coolant in nuclear power plants with VVER 440 reactor. Original nickel gasket is replaced by kammprofile gasket with expanded graphite. Calculation results are compared with each other and are described as suggestions for further calculation. The first part included a detailed description of the primary coolant water chemistry, along with a description of the function and structure of ion filter.
Heat and stress analysis exchanger
Jedlička, Rostislav ; Šen, Hugo (referee) ; Baláš, Marek (advisor)
The main goal of master’s thesis is a propsal of the heat exchanger. The heat exchanger is double way of water side with the integrated air-vapor mixture cooler. Another aims are heat computation, heat loss computation, solidity dimensioning and selection of a material for a major selection. The last task is about detection of a vapor pressure trought the tube bundle.

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