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

The design and use of passive safety systems as one of the effective ways to increase reliability and safety level of nuclear power plants in new generation reactors has received much attention. In this paper, using RELAP5 computational code, the thermo-hydraulic analysis of the Passive Heat Removal System (PHRS) in the VVER-1000 reactor is investigated to ensure the capability of its performance under various conditions. For this purpose, the system nodalization is first extracted according to the available documentation and then the RELAP5 code input is developed. The results show that for 6.27 MPa pressure in steam generators and air temperature of 36 ° C, the system is able to remove 95.52 MW heat.

The present study deals with a new method for measuring the power of a reactor. This method uses gamma and neutron radiation resulted from the entire reactor structure, without changing its structure (online). In terms of functionality, this method can measure the reactor power in real-time and report it instantly. In order to obtain the relationship between reactor power and gamma and neutron leakage radiation by simulation, the values of the F5 tally are calculated at different distances from the reactor wall, using the MCNP5 code (Monte Carlo N-Particles). Then, we plot the diagram to observe the variation trend. However, an increase in the distance from the reactor reduces both the amount of radiation and the energy of the radiation particles until the neutron leakage radiation reaches zero within five meters of the body, due to its good insulation. However, this number is only about four meters for gamma irradiation. The reactor power can be calculated by measuring the flux at the given point and any time by determining the point for a range of leakage radiation reduction to background and the linear relationship between power and leakage flux.

The loss of coolant accident is due to the reduction of the coolant fluid volume in the first circuit. The direct cause of this accident is the mechanical failure or fatigue in the material of the first circuit components during the power plant operation. This accident, which is a design-based accident, is an important factor in assessing a nuclear power plant safety. If the break occurs in the main circuit of the first circuit with a diameter greater than 25% of the cross-section area, it shall be referred to as a large break. In this paper, this accident with a break diameter of 850 mm is modeled and analyzed using the TRACE code in a VVER-1000 reactor. The TRACE code is specifically designed for coolant loss accidents. With the help of this analysis, it is possible to have an accurate estimate of the reactor's safety and to obtain significant economic considerations instead of conservative assumptions assessment. Finally, the results of the TRACE code have been compared with the final safety analysis report of the power plant as well as previous research by the RELAP5. The results indicate the accuracy of the TRACE code in modeling the large break accident.

In this paper, uncertainty and sensitivity of emergency cooling systems of the primary and secondary circuits are investigated in the VVER1000/V446 nuclear reactor during the small break
(SB)-LOCA. The emergency systems include accumulators, high pressure and low pressure injection systems and emergency feed-water system. The analysis is performed using the RELAP5/Mod 3.2 Code for nodalization and simulation of the nuclear power plant. The GRS method is used for the calculation of the minimum number of the code to achieve 95% of the tolerance limit and confidence level. For a conservative analysis of the accident, the loss of power to the NPP and the failure of the two diesel generators are considered when SB-LOCA occurs. These limitations cause the malfunctioning of the two channels of the emergency cooling systems. Also, by considering the single failure criteria, one accumulator is failed throughout the accident process. The results show that the accumulators deuelop the most sensitivity of the maximum clad temperature. Also, the HPIS pumps have relatively little influence on the SB-LOCA. When all emergency cooling setpoints are altered, a new set is achieved, which causes to lower the maximum clad temperature, in comparison with a default setpoint of the emergency cooling systems.

The fresh fuel for BNPP-1 is due to be transported inside special containers which are supposed to be designed to stand against vibrations and impacts in order to protect the fuel from any possible damage. In order to perform the structural dynamic analysis of the containers, while being subjected to impact of dynamic forces, it is necessary to perform the vibration analysis which will lead to the vibrational modes and their natural frequencies for the structure of the containers. The vibration opposition of the containers must be far beyond the critical resonance. The resonance frequencies about the natural frequency of the structure will cause the enhancement of the oscillation range and may be ended with its disintegration. Determination of the natural frequencies and their mode shapes can be achieved by vibration-analyzing-methods. The amount of the natural frequency of any structure depends strongly on its shape, material and its lean points, as well as the amount and the type of the loads which the structure will be subjected to. In the presente research, the container of the fresh fuel of BNPP-1 is simulated by ANSYS®10.0 and their ten natural frequency modes have been calculated.