نشریه فناوری و انرژی هسته ای
پیاپی 1 (بهار 1401)

Evaluating fuel performance in the event of an accident is an important challenge in safety analysis. To this end, several computational codes have been developed to evaluate the performance of the developed fuel rod. These codes examine the phenomena that governing the behaviour of the fuel rod during an accident include sheath phase alteration, zirconium oxidation, creep deformation, and sheath alloy failure. The aim of the present study is to investigate the performance of the VVER-1000 reactor fuel rod during an LOCA accident using FRAPCON-4.0 and FRAPTRAN-1.5 codes. The sample examined is the VVER-1000 reactor fuel rod in the IFA-650.6 test. Comparison of the obtained results with other published theoretical and experimental results confirms the validity of the code in thermo mechanical analysis of VVER fuel rod. In addition, the simulation results showed that the current safety criteria in the design of LOCA safety systems, it is necessary to consider sheath failure as another safety criterion. This is currently being addressed by the German Commission on Safety (RSK) in the analysis of the LOCA incident.

The accurate calculation of effective kinetics parameters is very important in the study of time-dependent behavior of neutron populations.To calculate the parameters, it is necessary to calculate the flux and adjoint flux in the phase space, which can be calculated with deterministic and probability methods. In this paper, in order to obtain the kinetic parameters of the new configuration of the Isfahan heavy water zero power reactor core in a deterministic approach, the MTR-PC package has been used and in the probabilistic approach the developed mcnp code with Iterated Fission Probability (IFP) has been used.The kinetic parameters calculated by these methods are validated by comparing the response of the point kinetics and experimental values for a reactive injection, and a good agreement was found between the results of the calculations and the experimental values.

In this work, the leakage current variation of a silicon diode, as the basic element of many electronic components, has been investigated in the exposure by neutron spectrum of a typical thermal rector. To determine the PKA spectrum, the MCNPX Monte Carlo code has been used to calculate the non-ionizing energy loss in the device. Molecular Dynamic Calculations are used to determine the number of reactor neutron induced defects in a silicon diode. The simulation of electrical parameters for irradiation of reactor neutron was also done by SILVACO software. The results show that the leakage current increases by about 6.82 times the amount of it before irradiation, up to about 3.54 nA/μm by the exposure of neutrons.

Fluoride Uranium-Thorium molten salts are widely used in the molten salt reactors in different compositions. The most important characteristic of the MSRs is high melting and boiling point. Therefore, the use of molten salt solution of uranium fluoride can eliminate the need for heat transfer in radioisotope production systems by accelerators. The initial study of the use of this molten salt as a target material in a cyclotron accelerator was carried out to generate various radioisotopes. We used MCNPX and ALICE/ASH. The results of this study show that the use of these molten salts can solve the problem of heat transfer and limit the application of proton beam flow on the target material. While also producing and harvesting various radioisotope products simultaneously.By optimizing the target geometry, we can increase the production rate of the different fission products in molten salt material. In addition, the use of accelerators can reduce the dependence of the production of radioisotopes used in medicine and industry to nuclear reactors.

Gas-liquid two-phase flow is probably the most important form of multiphase flows and is found widely in the oil industry. The accurate prediction of the air and water flow-rates are important in two-phase flow. Nowadays, multiphase flow-rates measurement by gamma-ray attenuation technique is known as one of the most common precise methods. In this work, the air and water flow-rates independent of flow regime changes were accurately predicted within a two-phase flow loop in the laboratory. For this purpose, a combination of single beam gamma-ray, single detector and artificial neural network (ANN) were used in order to predict the flow-rates in the bubble, plug, slug, annular and dispersed regimes of gas-liquid two-phase flows. Two different types of neural networks (GMDH) were developed. The networks were developed based on four features extracted from recorded pulse height distribution in a dynamic condition. The result shows, air, and water flow-rates were measured with an average of Mean Relative Error (MRE) less than 4.5%. Overall results revealed that using the proposed method, gamma-ray attenuation technique combined with an ANN model can be efficiently used to predict the flow-rates. Furthermore, in this study, a new method based on a single beam, single energy, and the single detector was proposed in order to solve this problem, without any recalibration

One of the most important aspects in design and operation of a nuclear reactor is investigation of the reactor during transient and non-steady conditions. For this purpose, different methods are presented for transient analysis. The direct solution of space-time dependent neutron diffusion equation with delayed neutron precursor equation is one of the most accurate and computationally expensive method. Use of this method in power control system design results in complex controller that its implementation encounters difficulties. Therefore, the point kinetic equation is used in design of nuclear reactor power control system by losing of the neutron flux shape variations. Recently, the multipoint kinetic equations are presented to decrease this limitation. This paper investigates the solution of multipoint kinetic in nuclear reactors and compares its results to conventional point kinetic equation and time-dependent solution of neutron diffusion equation for different benchmarks. The multipoint kinetic equations are derived via two group diffusion equations and are solved with forth order Runge-Kutta method. The results indicate that the deviation of the power level by use of multipoint kinetic method is lower than the conventional single point method in contrast to time-dependent solution of neutron diffusion equation. The deviation of power level is getting bigger in large reactivity insertion/withdraw that small reactivity perturbation. Therefore, multipoint kinetic method is more recommended for small reactivity change.