Volume 88, Issue 1

Safety analyses of nuclear power plant accidents are a key tool in demonstration of nuclear safety and reliability. The chain of the safety analyses comprises several follow-up activities like neutronic core calculation, primary system simulation and containment response to a loss-of-coolant accident. The analyses can be done with numerous computational tools like the GRS codes ATHLET and COCOSYS. The COCOSYS and its predecessor RALOC have been used in a variety of safety analyses at UJV conducted in the framework of technical support of the Czech nuclear power plants. Application of the code in the field of nuclear power plant safety analyses is ruled by the Czech legislative, which defines a process called standardization. The standardization procedure equivalent to validation and verification process provides further independent validation on the level of user organization. In past decades, COCOSYS has been successfully used in numerous DBA and DEC-A analyses of VVER-1000/V-320 containment, evaluating both thermal hydraulic conditions as well as fission product source term. This paper summarizes independent COCOSYS validation at UJV, including setup of the iodine dry paint deposition model to Ameron Amerlock paint. Furthermore, a VVER-1000/V-320 containment model is presented and analysis of DBA and DEC-A LB LOCA at VVER-1000/V-320 is conducted. The evaluation of the results aims at confirmation of acceptance criteria related to containment and impact of containment spray system. An emphasis is given to evaluation of containment source term, where comparison of both events is conducted for release of iodine, cesium, and xenon.

Nuclear Facility (NF), during shutdown and startup, are in the essential need for reliable electric power that should be delivered by electric power grid to NF. Safe operation of NF needs a limited variation in both frequency and voltage.The reduction of power losses, improving voltage profile, and frequency in electric grid connected with NF can be achieved by optimally distributed generators (DGs) placement. This paper presents a mathematical model for multible types of DGs placement in electric grid feeding NF. Also, it proposes artificial intelligence solution methodology for active and reactive power DGs placement problem. The trained Adaptive Neuro-Fuzzy Inference System (ANFIS) with Cat Swarm Optimization algorithm (CSO) is used for optimal solution. The optimization technique is tested and validated by using different sizes of electric grid. Test results showed a more reliable and efficient approach compared with other approachs.

The investigation of nuclear power plants (NPPs) accidents in the past shows that most of the accidents mainly occur in unexpected events. In this study, in order to verify whether Ecological Interface Design (EID) improves the situational awareness of operators in NPPs, this paper first analyzes the system by using the first three stages of cognitive work analysis, and then applies EID to the operating interface of NPPs to develop an ecological interface. In order to make the test results more complete, an improved interface has also been developed. A process expert and six operators were invited to participate in our experiments to measure situational awareness. The results show that the situational awareness of ecological interfaces in unexpected events is significantly higher than that of traditional and advanced interfaces. The significance of this study is that EID, as a practical technology, can be widely used in operator control rooms to improve the ability of operators to solve unexpected events.

Nowadays, reducing fuel consumption, fuel cost, and fuel waste is obtained by increasing the burnup and reactor cycle length as essential goals in nuclear power plants. Due to the scarcity of Iran’s uranium resources and the plan for using new types of fuel in the Water–Water Energetic Reactors (WWER-1000) at Bushehr Nuclear Power Plant (BNPP), utilizing Integral Burnable Absorbers (IBAs) is of great importance. In the present study, a neutronic evaluation has been conducted to investigate the variations of infinite multiplication factor versus burnup, reactivity swing, and power distribution in various IBAs, including Gd 2 O 3 -UO 2 , Er 2 O 3 -UO 2 , and Dy 2 O 3 -UO 2 . The results were compared with the standard burnable absorber in the BNPP reactor core (CrB 2 Al). It can be concluded that gadolinium IBA, with a concentration of 5%, has the greatest effect on the initial reactivity and reduces the reactivity swing by 19% compared to the CrB 2 Al BA. In addition, by using coaxial gadolinium-erbium IBA pins, as well as optimizing the neutronic condition, the reactor cycle length increased by 1.01 GWd/MTU compared to the standard Bushehr BA.

Lives are threatened whenever there is an act of theft or destruction against a nuclear and radiological facility thus, physical protection systems are effectively employed to prevent or mitigate loss of valuable assets. Health facilities that provide radiotherapy services have in their facility a temporary storage unit where disused radioactive sources 60 Co, 137 Cs and 192 Ir are kept and this raises concerns for effective physical protection. An integrated physical protection System was considered to consolidate of all sub systems, sensors and elements related to protection system for an effectively secured environment at a radiological facility. Sequence adversary diagram (ASD) was developed to depict the paths that enemies can take to achieve sabotage or stealing objectives and analyze flews in the paths. The approach to this security system effectiveness focuses on using probabilistic statistical methods for risk evaluation considering detection, delays and response. This paper considers the basic and effective elements required for physical protection system for a radiological center and makes risk evaluation as an approach of security system effectiveness which can serve as fundamental guiding principle for decision makers in the establishment of an effective physical protection for a radiological center.

MT-II is a spherical tokamak with a major radius of 0.15 m and a minor radius of 0.09 m, currently under development at the Pakistan Tokamak Plasma Research Institute. It is designed with a higher elongation of 2.67. This paper presents the design and material analysis of the limiter configuration for the MT-II tokamak, which is being carried out in two phases. In the first phase, theoretical studies and calculations are performed to estimate the plasma edge temperature, density, particle velocity, input power, heat flux, heat load and surface temperature on the limiter tile. In the second phase, computational techniques are applied to analyses the material properties, the maximum/minimum surface temperature rise (∆ T  °C) at stable heat load and power deposition based on theoretical calculations that will help optimize the design parameters of the limiter. The type of material and the surface temperature of the limiter as well as the general design parameters of MT-II are included in the proposed poloidal limiter. The results suggest that crystalline vein graphite is a suitable candidate for the proposed poloidal limiter. A combination of mechanical and electrical feedthrough techniques are used to improve the performance of the limiter. The proposed limiter is able to meet the requirements of MT-II.

In this paper a study of the behavior of Deuterium-Tritium (D-T) plasma nuclear fusion reaction in terms of variations of time and temperature and in the presence of deuterium-tritium sources using double cone ignition is presented. The aim is the determination of the optimum physical conditions with low tritium consumption rate for obtaining the total energy gain with a value of greater than 200.

The unattended ore grading radioactivity measurement system is an important part of the digital uranium mine. However, with the promotion of the digitization of mines, the radioactivity measurement system for the truck checkpoint used to measure the grade of ore for hydrometallurgical process still maintains the backward processing method of offline measurement and manual recording. To improve the situation, an unattended ore grading measurement system using gamma detectors is developed. Through unattended identification, weighting and radioactivity measurement, the values of the grade of ore and the uranium content are calculated and uploaded automatically. The system has been put into practical use at a uranium mine in southern China in September 2021. Through the field use, the system can realize automatic measurement, reduce labor costs, and establish a production database, which lays the foundation for further application in the construction of digital mines on a large scale.

Because of the prompt improvements in Micro-Electro-Mechanical Systems, thermal management necessities have altered paying attention to the compactness and high energy consumption of actual electronic devices in industry. In this study, 625 data sets obtained numerically according to the change of five different geometric parameters and Reynolds numbers for delta winglet type vortex generator pairs placed in a microchannel were utilized. Four dissimilar artificial neural network models were established to predict the heat transfer characteristics in a microchannel with innovatively oriented vortex generators in the literature. Friction factor, Nusselt number, and performance evaluation criteria were considered to explore the heat transfer characteristics. Different neuron numbers were determined in the hidden layer of each of the models in which the Levethenberg–Marquardt training algorithm was benefited as the training algorithm. The predicted values were checked against the target data and empirical correlations. The coefficient of determination values calculated for each machine learning model were found to be above 0.99. According to obtained results, the designed artificial neural networks can provide high prediction performance for each data set and have higher prediction accuracy compared to empirical correlations. All data predicted by machine learning models were collected within the range of ±3% deviation bands, whereas the majority of the estimated data by empirical correlations dispersed within ±20% ones. For that reason, a full evaluation of the estimation performance of artificial neural networks versus empirical correlations data is enabled to fill a gap in the literature as one of the uncommon works.

The average flow velocity in heat exchangers is considered less often and thus needs further and detailed investigation because of its crucial influence on the overall thermal performance of the application. The use of nanofluids has similar influences to finned tube designs. Considering the rise in heat transfer and pressure drop, uncertainties in cost analyses with the uses of fins and nanoparticles, evaluation of optimum operating velocity of the fluids is necessary. On the contrary, there aren’t enough experimental, parametric, or numerical investigations present on this subject. The use of machine learning techniques to heat transfer applications to make optimization becomes popular recently. In this work, important factors of the process as tube number, cleanliness factor, and overall cost as output factors have been estimated by an artificial intelligence method using 339 data points. The influence of input factors of Reynolds number, thermal conductivity, specific heat, viscosity, and total fin surface efficiency on the outputs have been studied. Total tube number, cleanliness factor, and total cost analysis have been determined with deviations of −0.66%, 0.001%, and 0.12% as a result of the solution with 6 inputs, correspondingly.

Production of high specific 60 Co activity in nuclear research reactor needs long time of irradiation to reach the proper consumer demands. Cobalt device is a fixed experiment inside the core of the second research reactor (ETRR-2) that designed for production of 60 Co with activity of 50,000 Ci. Cobalt device is assembly of 16 pins of aluminum cans that filling with the cobalt foils. Utilizing the programs that depend on the Monte Carlo method to estimate 60 Co activity is the objective of this study. MCNP6, KENO-VI, and MAVRIC codes would be used to evaluate the time that the cobalt device has been spent inside the reactor core for reaching the proper 60 Co activity considering the continuous irradiation as well as the discrete irradiation. The calculations show that nearly one year of continuous neutron irradiation is sufficient for reaching the required activity. In case of discrete irradiation, the time required for reaching the required activity would be more than a year depending on the reactor operation period.