|
|
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.
|
|
|
Sodium - Carbon-dioxide Heat Exchangers for Sodium Cooled Fast Reactor NPP (SFR)
Foral, Štěpán ; Šimo, Tomáš (referee) ; Matal, Oldřich (advisor)
This master’s thesis deals with a design of Na-CO2 heat exchanger. There is a comparison of shell and tube heat exchanger with PCHE in the first part. Further the shell and tube heat exchanger with internally finned tubes was chosen as the basic conception. There was performed an optimization of construct and operations parameters for this concept. The optimization was performed on the basis of thermal and hydraulic calculations. Further there were performed calculations for ensuring of safe operation of the heat exchanger. The conclusion of the diploma thesis deals with comparison of the designed heat exchanger with similar projects.
|
|
|
Sodium cooled fast reactors
Daňhel, Aleš ; Katovský, Karel (referee) ; Foral, Štěpán (advisor)
This bachelor’s thesis deals with the sodium-cooled fast reactors. It comprehensively describes the problem of sodium-cooled fast reactors. Attention was paid to the basic specifications and parameters of these reactors. There was briefly described nuclear reactions which are under way in core of sodium-cooled fast reactors but also chemical reactions which are linked to function of liquid sodium as a coolant and heat transfer substance. Attention was also paid to the differences in the reactor core configuration and to the machine device specific for sodium-cooled fast reactors. Further on this bachelor’s thesis puts on the overview of sodium-cooled fast reactors that have worked, are still working or are under construction in each country all over the world. There was briefly described generation IV nuclear reactors, particularly its history and reasons that originate generation IV. Under generation IV was also described nuclear reactor SFR and highlighted differences compared to existing sodium-cooled fast reactors. In the practical part of this bachelor’s thesis is easily made a calculation of the heat transfer from the fuel rod to the cooling sodium and there is also mentioned course of the heat transfer coefficient along the fuel rod. This calculation was made by computing program MATLAB.
|
|
|
Sodium cooled fast reactors
Daňhel, Aleš ; Katovský, Karel (referee) ; Foral, Štěpán (advisor)
This bachelor’s thesis deals with the sodium-cooled fast reactors. It comprehensively describes the problem of sodium-cooled fast reactors. Attention was paid to the basic specifications and parameters of these reactors. There was briefly described nuclear reactions which are under way in core of sodium-cooled fast reactors but also chemical reactions which are linked to function of liquid sodium as a coolant and heat transfer substance. Attention was also paid to the differences in the reactor core configuration and to the machine device specific for sodium-cooled fast reactors. Further on this bachelor’s thesis puts on the overview of sodium-cooled fast reactors that have worked, are still working or are under construction in each country all over the world. There was briefly described generation IV nuclear reactors, particularly its history and reasons that originate generation IV. Under generation IV was also described nuclear reactor SFR and highlighted differences compared to existing sodium-cooled fast reactors. In the practical part of this bachelor’s thesis is easily made a calculation of the heat transfer from the fuel rod to the cooling sodium and there is also mentioned course of the heat transfer coefficient along the fuel rod. This calculation was made by computing program MATLAB.
|
|
|
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.
|
|
|
Sodium - Carbon-dioxide Heat Exchangers for Sodium Cooled Fast Reactor NPP (SFR)
Foral, Štěpán ; Šimo, Tomáš (referee) ; Matal, Oldřich (advisor)
This master’s thesis deals with a design of Na-CO2 heat exchanger. There is a comparison of shell and tube heat exchanger with PCHE in the first part. Further the shell and tube heat exchanger with internally finned tubes was chosen as the basic conception. There was performed an optimization of construct and operations parameters for this concept. The optimization was performed on the basis of thermal and hydraulic calculations. Further there were performed calculations for ensuring of safe operation of the heat exchanger. The conclusion of the diploma thesis deals with comparison of the designed heat exchanger with similar projects.
|
guest ::
