Sharif Digital Repository

Fire is one of the most important events in nuclear power plants which has a large share of total core damage frequency. For example, the core damage frequency in Bushehr nuclear power plant from fire accidents is 6.1E-6/ year which is equivalent to about 63 percent of core damage frequency from all internal accidents. However, due to the unavoidable presence of flammable materials, e.g. Electrical equipment, cables, gas, flammable liquids, …), it is impossible to prevent fires in these plants . Fire at nuclear power plants may be caused directly or indirectly by one or more initiating events, which the following accidents in turn would threaten the safety of the plant. In this project,... 

Station Blackout (SBO) involves Loss of Offsite Power (LOOP) concurrent with failure of the onsiteemergency ac power system. Although SBO is a beyond designbasis accident, in recent years this accident attracts a lot of attention. Fukushima nuclear accident which was initiated by earthquake followedby tsunami led to a SBO accident. In this thesis, Probabilistic Safety Assessment (PSA) is applied to analyze SBO in IR-360 nuclear power plant and to compute its frequency. Meanwhile, weak points of the plant are identified in order to minimize the SBOprobability of occurrence. Fault tree method is used to evaluate emergency power system and offsite power system reliabilities. Common Cause... 

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... 

Transfer of Spent Nuclear Fuel from reactor building to storage facilities is one of the most important operations at Front End of Nuclear Fuel Cycle. The aim of this project is to calculate the risk of the radioactive materials release into the environment during various stages of spent fuel transfer operations. In addition, the contribution of each of the initiating events in total release probability were also determined. For this purpose, Probabilistic Risk Assessment method which is usually used in safety evaluation of the nuclear facilities, is applied. The main tools used in this method are event tree and fault tree. In this study, SAPHIRE software has been used to construct the event... 

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... 

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... 

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... 

The safety and optimal performance of nuclear reactors require online monitoring in the core. One of the most important requirements of core monitoring is the knowledge at all time of the neutron flux distribution in the core. The present M.Sc thesis describes a method which avoids the solution of time dependent neutron diffusion equation and uses online readings of the fixed in-core neutron detectors to reconstruct the three-dimensional (3D) neutron flux distribution. The essential idea of nodal synthesis method is separation of time and space dependence of the neutron flux distribution. The time dependent section of the flux distribution is determined by neutron detector readings and space... 

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... 

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... 

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... 

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... 

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... 

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... 

operation of a nuclear power plant. Therefore, constant monitoring of the reactor core with reliable methods is important. To monitor the reactor heart, it is necessary to estimate and calculate some parameters, with high speed and accuracy, such as power distribution inside the heart, reactivity feedback coefficients, PPF, DNBR, etc. Analytical methods are often used to calculate these parameters, which in case of failure of the sensors, the calculations will be practically disrupted, and the method used in this research can solve these problems by losing a small amount of accuracy.In this study using real data of Bushehr nuclear power plant (BNPP) and by soft computing methods and... 

Given many computing errors in current systems, a method appears necessary for predicting the nuclear parameter quickly and accurately. In this thesis, a neural network was used to predict safety in a nuclear power plant in order to develop an operating aid tool for preventive measures.First, some studies were conducted on appropriate feature selection for training neural networks. Some case studies have also been carried out on parameter prediction through soft computing in a power plant. In the next section, an expert judgment was taken into account to select DNBR (Departure from Nucleate Boiling Ratio) as a criterion for safety evaluation in the exploitation of a nuclear power plant (PWR)... 

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... 

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 the present work storage of hydrogen molecules in hybrid nanostructures has been evaluated. Hybrid nanostructures consist of carbon structure bases which have been doped by metal atoms. Carbon structures used in this thesis include graphene and metal organic frameworks (MOFs). Carbon structures have superior properties such as very low density, while metal atoms are considered to enhance the interactions and increase the hydrogen storage capacity. In the first step of the work, Monte Carlo method was applied to model the system. To develop the Monte Carlo method for hydrogen adsorption on graphene sheets, Feynman-Hibbs corrections were added to Lennard-Jones potential. However in the next... 

Because of the Complexity in Calculating the Probability of Failure of the System, This Computation is Done by the Software Have Been Developed for this Purpose. Of Course, When We Use Them, We See That Each of Them Have Their Advantages and Disadvantages. One of the disadvantages that have been tried to be resolved over time, the type of programming and algorithm for calculation of it is the fault tree Because new techniques like modularization which have been designed in recent years, a significant improvement in accuracy and speed performance computing applications that use this method have been obtained.In This Thesis Has Developed Software That is Able to Calculate the Probability of...