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Modification of TTL-1 Thermo hydraulic Test Loop for Study of SB-LOCA in Low Pressure and Temperature Conditions
,
M.Sc. Thesis
Sharif University of Technology
;
Vosoughi, Naser
(Supervisor)
;
Jafari, Jalil
(Supervisor)
Abstract
In this thesis, TTL-1 thermo hydraulic test loop which was constructed in the Nuclear Science and Technology Research Institute (NSTRI) has been modified. These modifications include changing core section from one fuel rod to a fuel assembly with triangular arrangement, adding a pressurizer on the hot leg of loop, an accumulator, LOCA valve and installation of new sensors for upgrading of data acquisition system and modification of loop piping design. By these modifications, we will able to study different accidents such as LOHA, LOFA and LOCA and validate nuclear thermo hydraulic codes. Moreover, this new thermo hydraulic test loop (TTL-2) has been modeled by RELAP5/MOD3.2 in order to...
Evaluation of Natural Circulation in Spent Fuel Pool of Bushehr NPP, in Case of Loss of SFP Cooling System
, M.Sc. Thesis Sharif University of Technology ; Ghofrani, Mohammad Bagher (Supervisor)
Abstract
The Fukushima accident is one of the reasons why the spent fuel pool (SFP) is an important safety element. The large amount of spent fuel inside the spent fuel pool has made it a source of hydrogen and radioactive material in the accident of any disturbance to pool cooling. Although the loss of spent fuel pool cooling system accident has a slow transient, it can lead to disasters if you ignore it. In this project, spent fuel pool of Bushehr Power Plant unit ‘I’ was simulated by Ansys Fluent and the natural circulation of fluids in the absence of cooling system was investigated. The distribution of water temperature in the pool was investigated for the worst possible spent fuel pool loading...
Thermal-Hydraulic Simulation and Analysis of Two-Phase Thermal Shock in Pressurized Light Water Power Plants
, Ph.D. Dissertation Sharif University of Technology ; Ghofrani, Mohammad Bagher (Supervisor)
Abstract
As a result of fission reaction in a nuclear reactor, the produced high neutron flux would affect the material of Reactor Pressure Vessel (RPV). This neutron radiation has a detrimental impact on the mechanical properties of the RPV material such as hardening (or embrittlement) while neutrons are absorbed by the material. A major concern in embrittled RPVs is propagation of critical flaw causing through-wall cracks. Some transients leading to overcooling of RPV intensify the propagation of theses cracks and result in thermal load on RPV, known as Pressurized Thermal Shock (PTS). Such situation could be created in case of Emergency Core Cooling System (ECCS) actuation which leads to injection...