|
|
Evaluation of Nuclear Fuel Safety and Reliability Using Research Reactors' In-Core Experiments
Matocha, Vítězslav ; Foral, Štěpán (referee) ; Katovský, Karel (advisor)
The aim of this master thesis is to show a connection among nuclear fuel safety, experiments led in research reactors and calculation codes. This thesis focuses on the calculation code Transuranus. There are represented four experiments, which were calculated in Transuranus. The fission gas release, elongation and growth of fuel were particularly monitored. Is is possible to set differences among versions v1m1j09 and v1m3j12 from achieved results, as well as the influence of selected Transuranus parameters on the results, so the thesis may bring new pieces of knowledge for improvement of safety analysis calculation by Transuranus.
|
|
|
Power and Temperature Distribution in Nuclear Fuel Assemblies of VVER-440 reactor at Dukovany NPP
Smola, Luděk ; Novotný, Filip (referee) ; Katovský, Karel (advisor)
This Master’s thesis focuses on calculation of power and temperature distribution in fuel assemblies of VVER-440 reactor at Dukovany Nuclear Power Plant. Theoretical section contains a brief description of VVER-440 technology, fuel and its development at Dukovany Nuclear Power Plant, basics of heat generation in nuclear reactors as well as an overview and categorization of computer codes, used for core calculations. Of these codes, the MOBY-DICK computer code is then described in depth, including its input and output files. The MOBY-DICK code is later on used for pinwise calculating power distribution of selected fuel cycles of defined units at Dukovany Nuclear Power Plant, with vizualization of output values for characteristic fuel assemblies. Results of this computation are then used for analysis, whether uneven power distribution in the core and heat generation gradient within fuel assemblies have any influence on measuring channel output temperatures, which is the pivotal part of this thesis.
|
|
|
Boltzmann equation eigenvalue calculation for nuclear reactor cores, spent fuel pools, and storage casks
Bílý, Lukáš ; Vojáčková, Jitka (referee) ; Katovský, Karel (advisor)
In my thesis, I focus on the computational tool SCALE, which utilizes computational codes for modeling and calculating critical values of reactors. Prior to the practical section, there is a theoretical part that focuses on reactor physics and computational codes. The selected computational code is the KENO V.a code, which employs the Monte Carlo method for calculating the effective multiplication coefficient. The computational code is used for modeling selected benchmarks from the ICSBEP database. At the end of the thesis, the average percentage deviation will be determined.
|
|
|
Power and Temperature Distribution in Nuclear Fuel Assemblies of VVER-440 reactor at Dukovany NPP
Smola, Luděk ; Novotný, Filip (referee) ; Katovský, Karel (advisor)
This Master’s thesis focuses on calculation of power and temperature distribution in fuel assemblies of VVER-440 reactor at Dukovany Nuclear Power Plant. Theoretical section contains a brief description of VVER-440 technology, fuel and its development at Dukovany Nuclear Power Plant, basics of heat generation in nuclear reactors as well as an overview and categorization of computer codes, used for core calculations. Of these codes, the MOBY-DICK computer code is then described in depth, including its input and output files. The MOBY-DICK code is later on used for pinwise calculating power distribution of selected fuel cycles of defined units at Dukovany Nuclear Power Plant, with vizualization of output values for characteristic fuel assemblies. Results of this computation are then used for analysis, whether uneven power distribution in the core and heat generation gradient within fuel assemblies have any influence on measuring channel output temperatures, which is the pivotal part of this thesis.
|
|
|
Evaluation of Nuclear Fuel Safety and Reliability Using Research Reactors' In-Core Experiments
Matocha, Vítězslav ; Foral, Štěpán (referee) ; Katovský, Karel (advisor)
The aim of this master thesis is to show a connection among nuclear fuel safety, experiments led in research reactors and calculation codes. This thesis focuses on the calculation code Transuranus. There are represented four experiments, which were calculated in Transuranus. The fission gas release, elongation and growth of fuel were particularly monitored. Is is possible to set differences among versions v1m1j09 and v1m3j12 from achieved results, as well as the influence of selected Transuranus parameters on the results, so the thesis may bring new pieces of knowledge for improvement of safety analysis calculation by Transuranus.
|
guest ::
