Sharif Digital Repository

The capability of Bushehr nuclear power plant spray system to reduce fission product and control a specified severe accident has been investigated in this study. In order to evaluate spray system in one of the worst cases for containment, we have chosen Large Break Loss of Coolant Accident (LB-LOCA) without access to active part of ECCS system. We have only simulated containment, so the leakage of various materials from core and primary circuit of nuclear power plant are considered as the boundary conditions for simulation. Following the Fukushima accident in 2011, many nuclear safety legislative groups concentrated on using mobile equipment which is essential for nuclear power plants these... 

During the operation of a nuclear power plant, the reactor pressure vessel (RPV) is exposed to a variety of pressure and thermal stresses and neutron radiations. This will cause the loss of the initial strength of the reactor vessel component. During the occurrence of some of accidents, an excessive cooling of the coolant inside the RPV takes place which in the term is called overcooling transients. In addition, in some of these events, a break in a section of the circuit will reduce the water level at the core of the reactor. By reducing the water level, the existing emergency makeup water systems are activated and inject water into the reactor. The temperature of the added water is much... 

In this study, the determination of the distribution and the differences concentration of hydrogen in containment of Bushehr power plant by modeling of containment and Hydrogen removal system of Bushehr power plant during LB-LOCA accident is accomplished. In Bushehr nuclear power plant, which has a unique design, the containment is similar to PWR power plants with a metal sphere, but the installed reactor is a VVER-1000. After a full study on the containment and its systems, the required data for modeling is gathered and afterwards, the engineering handbook is prepared for MELCOR input code and then the modeling and size classification of the containment is done in 4 different ways,... 

In this work the large break loss of coolant accident at Boushehr Nuclear Power Plant (NPP) has been studied. At first, primary and secondary side’s components which play important role in the accident have been modeled using RELAP5/MOD3.2 GAMMA system code. The components are: reactor pressure vessel, main pipe lines, main coolant pumps, pressurizer, steam generators, steam lines, emergency core cooling accumulators, boron solution supply tanks and pumps (LP& HP) and KWU accumulators. After preparing the input file and running the code, thermal hydraulic properties such as radial distribution of temperature in fuel rods, temperature distribution in steam generator primary and secondary... 

In this survey the TTL-1 thermal hydraulic test loop in Nuclear Science and Technology Research Institute (NSTRI) has been modified to study thermal hydraulic phenomenon in research reactor thermal condition. These modifications include: replacing the test section containing a single heating rod by a group of 31 rods in a triangular array, relocating the pressurizer from the cold leg to the hot leg and improving piping of the test loop. In order to measure the time dependant flowrate of the coolant, a rotameter flowmeter is added. Three level meter measure the water level in pressurizer, test section and the primary loop drain tank and a 3-phase AC inverter controls the rotational speed of...