نشریه فناوری و انرژی هسته ای
سال دوم شماره 1 (پیاپی 5، بهار 1402)

There are different parameters such as increase or decrease of heat removal from primary circuit which affect the coolant mixing phenomena in the vessels of pressurized water reactors (PWRs). Determination of mixing level is very important from reactor safety and control aspects. In this study, the thermal hydraulic test of coolant mixing within the reactor pressure vessel of Bushehr nuclear power plant (BNPP) has been simulated (3-D modeling) using ANSYS CFX 18.0. In this test, the fluid mixing due to primary circuit heat removal decrease has been investigated and the goal of this research has been defined as finding the coolant temperature distribution, computing the primary circuit loops mixing coefficients and other thermal hydraulic parameters of coolant in the whole reactor zones specially the reactor core which is the most important. To achieve this, the geometry of whole reactor considering all components have been modeled and the governing equations of reactor flow field (the Reynolds Averaged Navier-Stokes equations utilizing SST k-ω turbulence model) have been solved in CFX. Comparison of simulation results and experimental results of BNPP startup test shows the average error of 6.45 % and 10.92 % for mixing coefficient of loops and core inlet, respectively. According to the implemented simplifications, the results have good accuracy.

Due to the many advantages of small modular reactors, including high safety, easier commercialization, and higher efficiency, are the design, development and construction plans of many countries, especially the leading countries in the nuclear industry in the Stable, safe and economical energy production program. Hence, in this paper, an attempt has been made to analyze the role of reflectors in determining the control and neutronic parameters of the CAREM-25 advanced small modular reactor core using the code MCNPX2.7.0. code modeling. In this regard, the reactor core was modeled in two states of presence and absence of reflector to calculate and compare the K-factor, excess reactivity and neutron flux distribution in the both two states. Results illustrated that a significant volume of water as a reflector had a large effect on the neutron population of the core of this reactor and could increase the excess reactivity by about 20 (mK).

The double-spherical harmonics method (DPN) is a common approximation in the study of the neutron transport equation in reactor physics problems. Inside a reactor near points where strong discontinuities in material properties occur, such as bare boundaries or areas near strong absorbers, there is usually greater anisotropy in the angular distribution of neutron flux. A more appropriate description of the angular flux behavior at such points requires the use of methods such as DPN, which utilize separate expansions for different directions of neutron motion instead of using a single expansion for all directions, as in the PN method. In this paper, the multigroup, one-dimensional neutron transport equation in the Cartesian coordinate system is solved using the DP1 approximation. To do this, first the multigroup DP1 equations and the corresponding boundary conditions are derived, then they are written in the form of multigroup neutron diffusion equations, which are here called the simplified-DP1 or SDP1 equations. The finite element method is then used to numerically solve the SDP1 equations. The results of the proposed method are discussed for several different test problems in comparison with the P3 method.

Safety, reliability and maintenance are key issues in the operation of nuclear power plants. Proper and organized implementation of maintenance activities can provide a platform for the safe operation of the power plant.In This research we try to increase the reliability of power plant systems and equipment by optimizing maintenance activities by introducing the reliability-based maintenance method (RCM).In This article the different approaches in the RCM program implemented in nuclear power plants and tries to choose an operational and efficient method in Bushehr nuclear power plant. Hence, the conflicts between classical RCM and simple RCM have been raised and an attempt has been made to resolve these conflicts through a comprehensive model.In this paper, two different RCM methods based on engineering judgment and statistical analysis have been used and the effect of these methods on heart rate has been evaluated. The method of combining FTA and recursive FMEA to evaluate equipment and select critical equipment according to the depth of analysis Optimally, used. In this method, by performing FTA and FMEA repeatedly at different levels of power plant analysis, critical equipment is identified using probabilistic safety assessment criteria

In this research, a mathematical model for dynamic behavior of a small modular reactor core with natural circulation flow is presented by Simulink toolbox of Matlab software. First, point kinetic equation by considering one group delay neutron simulated in the Simulink environment to investigate the time dependent neutronics behavior that important parameters of this equation such as mean generation time, delay neutron, fuel and moderator reactivity coefficient are calculated using the MCNPX code. Mann's model with one-dimensional Dittus-Boelter heat transfer coefficient was also used to investigate the heat transfer process from fuel to coolant, and after completing the proposed model, the system's response to insert of a step reactivity was evaluated. The results of this research show that the proposed model has ability to predict the dynamic behavior of a small modular reactor core with natural circulation flow.

The main purpose of the present study is the evaluation of the candidate sites for proposed Fars research reactor. Because the proposed research reactor is being designed conceptually and is very similar to the Tehran Research Reactor, the specifications of the Tehran Research Reactor have been used in this study. In this research with the help of site selection criteria and geographic information system utilities in Arc-Gis 10.4.1 software for three regions of Shiraz, Marvdasht and Sarvestan. The results show that Bajgah site in Shiraz region is the preferred site for Fars research reactor. The process of present research can be used as a methodology for site selection of other similar nuclear facilities.