Sharif Digital Repository

Various factors, such as increasing or decreasing the heat removal from the initial circuit, or increasing the flow rate of the cooling fluid in the reactor, causes the phenomenon of the coolant mixing in the PWR reactors. In this project, the thermohydraulic test of coolant mixing has been simulated in the pressure vessel of the Bushehr nuclear reactor. In this test, the mixing of the coolant caused by the reduction of heat removal from the primary circuit by the secondary circuit is investigated. In this case, the primary circuit temperature increases in the loop where the heat removal is reduced. The most important consequence of this event is the reactivity changes at the core of the... 

In order to ensure the safety of the plants, a set of commissioning tests based on international standards for nuclear power plants is carried out before operation. In this research, one of the Bushehr power plant commissioning tests, the loop connection primary circuit cooling pump test, was simulated using the RELAP5 code. In this test, the effects of loop connection primary circuit pumps on the thermohydraulic parameters and the eneterance of test-related systems was evaluated. After simulating the systems related to the test and extracting the results of the stable state, were evaluated the transient results was obtained with the experimental results of the test of loop connection the... 

In the absence of appropriate control, several accidents in the nuclear power, lead to serious consequences including core melted and radiation leakage outside reactor containment. In all these events, when core melted, the reactor operation is always finished and all efforts are then directed to environmental impacts of radioactive materials to the external environment. However, the occurrence of events that would bypass reactor containment, have special significance, because all possible efforts in order to limit the leakage of radioactive materials to the external environment are ineffective. In this respect the occurrence of such incidents, as most of which is considered the worst... 

Efforts to model and control surge, as the underlying instability of centrifugal compressors, were conducted. Of the three major instabilities in compressors, namely chokage, stall and surge, the latter has a significant role in limiting the available range of operation in centrifugal compressors, which are the most common devices for the transportation of Natural gas. The customary method to circumvent surge, which if occurred, could impose severe, catastrophic and irreversible damages to the machine, is to avoid the zones in which it is likely to develop by means of a recycling system. These often called "surge avoidance" or "surge prevention" schemes have the benefit of high reliability,... 

LOOP accident is one of the most important accidents in NPPS and it leads to loss of power supply of the RCP pumps which are the main pumps of primary circuit. Subsequently trip of the rector occurs. In safety analysis, concerning worst conditions, assumes that after a small period of time, main generator of plant stops working and loss of AC power supply of all power plant loads including safety systems ocures. To control of this accident, it’s necessary to startup and load backup power supply system (EDGs) for at least one safety channels, if not, the accident moves forwards and becomes S.B.O accident which is more important accident that ends to core damage because of stops working of all... 

This thesis introduces a new method for identification and forecasting of future states of nuclear power plants (NPPs) parameters. The proposed method consists of four steps. First, the type of transients is recognized by the modular identifier which has been developed using the latest advances of error back propagation (EBP) learning algorithm. In second step, for more robustness of modular identifier against noisy input data, auto-regressive integrated moving average (ARIMA) method is used. A hybrid network is then used to forecast the selected parameters of the identified transient. ARIMA model is used to estimate the linear component of the selected parameters. The neural network... 

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,... 

Due to nuclear power plants’ growing roll in providing energy, analyzing probable accidents and their prevention is of vital importance. The thermal-hydraulic shock accident which happened in Bushehr nuclear power plant deaerator has been simulated in this study. Steam bubble collapsing which takes place in two phase medium by mixing steam and subcooled water leads to this kind of shock causing financial and physical irreparable damage in nuclear power plants. For carrying out this simulation, first an steady state model of the deaerator was run using RELAP5 code and then using this model and other plant’s data such as the ropert of the accident and plant’s Final Safety Analysis Report... 

Safety is an important factor in the nuclear power plants due to its inherent human and environmental hazards. Therefore, in order to ensure the safety of the plant and proper functioning of its various components, a series of commissioning tests should be conducted according to international nuclear power plants standards, before its operational exploitation. On the other hand, safety of the plant can be guaranteed by modeling transient states and possible scenarios of accidents in different situations. In this study, a commissioning test of Bushehr Nuclear Power Plant (BNPP) and its associated accident is modelled by gathering geometric data and information about the facilities... 

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... 

Following Three Mile Island (TMI) accident in 1979, first severe accident (SA) in Nuclear Power Plants (NPPs), Accident Management Support Tools (AMSTs) were developed and installed in a number of NPPs. Lessons learned from Fukushima accident highlighted importance of Accident Management (AM) in mitigation severe radiological consequences after a SA and suggested reconsiderations of AM program which in turn created the need for AMSTs adaption and modernization. An efficient AMSTs should have the following principal capabilities: (1) Identification of accidents and diagnosis of the plant damage state (PDS), (2) Prediction of accident progress path and (3) Source term analysis and prediction... 

Identification of different hazards in a UF6 production process, and evaluation of the risk originated from these hazards is the main objective of this project. A number of hazards are present in a typical UF6 production process, such as leakage of chemical gases like HF and F2 and also radioactive UF6 gas release. In order to evaluate risk due to these hazards, probabilistic approach has been used. Due to lack of probabilistic safety criteria (PSC) for chemical releases, only for UF6 gas release risk assessment has been done. As a first step in PSA of this process eight groups of initiating events have been identified using HAZOP study, and for each initiating event, event tree analysis... 

Loose parts in the primary circuit of a nuclear power plant, causing damage to the fuel rods and other equipment, so early localization and mass estimation of this pieces can provide context of safety measures for the reactor after an event.Acoustic signals emitted by the location of these parts provide enough information to estimate their location and mass. So with obtain time-of-arrival differences between sensors and sound velocity can be estimate loose part location. In this thesis signals of the sensors in Bushehr power plant monitoring system are analyzied. To estimate the loose part locations, the time delays between sensors must be calculated. The time difference between the sensors... 

In this study the modeling of Bushehr power plant containment is considered. In Bushehr atomic 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 date 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, including the control volumes of 1, 9, 23 and 30 and then the temperature, pressure and density of hydrogen is examined. The result of studying control volume sensitivity reveals that 23... 

Inadvertent opening safety valve of pressurizer (SVP) is one of the most common initiator events in Booshehr nuclear reactors with occurrence frequency of 1.1E-0.2. This event has the third rank in initiator event among 16 other events (Loop and Compensated LOCA were ranked as first and second respectively). So special attention must be paid to SVP and the operators need to have specific trainings on this matter. In this thesis, the SVP as initiator event is being considered and the sequence which lead to core damage (CD), according to design informations is being modeled moreover the qualitative and quantitative analysis is carried out. In the modeling process we have used the SAPHIRE... 

In general, release of hazardous gas in chemical process plants could be due to two reasons. First is failures due to degradation of components that have inventory of hazardous gas in standard process conditions, and the second is deviation of system from standard process conditions because of different failure modes of related components, e.g. controller equipments. The purpose of this study is, calculating the probability of H2S releasing in GS(03) plant of Arak Heavy Water factory, using PSA methods and its risk assessment. At First, fifteen groups of initiating events were identified with Hazop study for desired system. Then, for calculating the frequency of each event, the combination... 

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... 

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... 

In this study, the thermal-hydraulic analysis of prismatic HTGR’s core with natural convection has been studied using porous media approach. VHTR are the new generation reactors which due to special neutron and thermo physical properties have highly inherent safety. In lose of forced circulation accident, decay heat is transferred from core to pressure vessel wall and then to water tubes in concrete wall at reactor cavity with conduction, convection and radiation automatically. In this case the high volume of decay heat is stored in graphite block with high thermal capacity and is prevented the instantaneous temperature rising.
The aim of this study is justifying inherent safety of HTGR... 

HTGRs are considered as 4th generation reactors which have prominent characteristics such as inherent safety, lower safety costs, High efficiency and high temperature applications. the most important challenges in developing these reactors is providing appropriate codes in design, simulating their performance and analysis of them. In this thesis, a japanese prismatic HTTR reactor has been selected as a reference reactor and the neurotic’s calculations implementation studied at cold zero power (CZP) and hot full power (HFP) states using DRAGON cell calculation codes and DONJON core computations. At CZP state, One group and two group radial & axial flux distribution, control rod critical...