جستجوی مقالات مرتبط با کلیدواژه "vver-1000" در نشریات گروه "فیزیک"

Event tree analysis is applied to quantify the core damage frequency (CDF) and assess the risk of nuclear power plants (NPPs) resulting from various postulated initiating events. To calculate this criterion, it is necessary to generate the probable scenarios according to the function of safety systems and the operator's actions. The classical event tree is currently used in PSA analysis. This method does not consider the accident's dynamics and scenarios. It considers only the availability/unavailability of the safety system functions and the operator's actions to calculate the frequency of each scenario. In contrast, the dynamic event tree method applies physical and probabilistic models to generate branches in the event tree, calculate the frequency of each scenario, determine the time profile of core damage, and time variation of physical parameters of the NPP for each scenario. This paper develops the dynamic event tree for the SBO accident at the VVER-1000/V446 NPP using the RELAP5 and RAVEN codes. The results are then compared with the outputs of the classical event tree. The results show that according to the assumptions, 3170 scenarios are evaluated in the dynamic event tree, while only 33 predetermined scenarios are examined in the conventional event tree. The calculated core damage frequencies are 3.61×10-6 (yr-1) and 1.97×10-6 (yr-1) for conventional and dynamic event trees, respectively.

The displacement of the atoms from their lattice sites is one of the results of neutron irradiation on materials. The primary knocked-on atoms spectrum and their angular distribution should be calculated for consideration of neutron radiation damage calculation. The AMTRACK program has been developed to calculate this information. This program extracts and analyzes information about the primary knocked-on atoms by using Ptrac output of the MCNPX code. In this study, Stainless Steel 316, one of the most important alloys in the reactor pressure vessel, is investigated. The material is irradiated by single energy neutrons of 1keV to 10MeV, the fraction of produced PKAs, their average energy, their maximum energy, and radiation damage value are calculated. The calculations are performed by using the neutron spectrum of the Bushehr reactor. Using this method, the amount of damage in the Bushehr reactor pressure vessel (for Stainless Steel 316) is equal to 7.2×10-22 (dpa/ fluence).

The article aims in the studying of thermal-hydraulic simulation of the VVER-1000 reactor core fuel assemblies’ reaction to the mass flux changes which are caused by the lose of coolant accident and its sudden pressure drop. The analysis of mentioned accident is performed in concise periods (mili second) by the use of the sound effect. Time-related thermal-hydraulic equations were analyzed by the method of a compressible fluid in a single heated channel and were evaluated by the results of the mentioned transient, in a PWR reactor. The mentioned transient was simulated in RELAP5 code and results were compared to the previous ones. Then, 28 reactor fuel assemblies were studied, considering the 1/6 symmetry of VVER-1000 reactor and unique features of every assembly. Mass flux drop was happened the end of the channel, after a few seconds. It was observed that mass flux is at dependent on the role of every assembly in the production of core heat power. The acoustic effect reveals some of the perturbations in mass flux changes, considering every fuel assembly features.

The main purpose of the work is to develop a valid physical model and an accurate numerical technique to describe the occurred phenomenon of the VVER-1000 fuel rod, during its lifetime, especially for high burn up conditions. There are many factors involved in the fuel rod performance, which each of them, intricately affect its behavior during normal operation. The accurate prediction of the fuel behavior is obligatory and will be utilizable for the fuel designers. In geneval, fuel rod behavior is affected by the various chemical, mechanical, and thermo neutronic phenomena. For a detailed assessment of the fuel behavior inside the reactor core, the mentioned factors and the dominant aspects must be modeled accurately. The physical models and correlations used in this paper, are chosen in such a way that all the simulation results be in a good agreement with the available post-irradiation examination data (PIE) and outputs of the FRAPCON-3.3 fuel rod performance code.