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BackgroundInterfacial area concentration (IAC) is a key parameter of the interface transfer term in the closed two-fluid model of two-phase flow, which characterizes the strength of the gas-liquid interface transport capacity. There are usually some methods for modeling and predicting the interface area concentration, such as empirical correlation formula and interface area transport equation, but these methods have large data dependence.PurposeThis study aims to provide direction for model revision and improve the prediction accuracy of IAC by adding interpretability to the neural network model.MethodsThe prediction model of IAC based on a neural network was firstly established for better prediction of IAC with two-phase flow. Then, different bubble behavior, physical relationships, and statistical distribution were combined, and the predictive ability of the neural network model with different input feature combinations was compared and analyzed by the post-interpretability method. Finally, based on the structure parameter size of each layer of the neural network, the appropriate data preprocessing method was selected by analyzing the output proportion.ResultsThe post explanatory analysis show that the maximum prediction accuracy of the neural network reaches 95.62% when the inputs of the neural network are the gas superficial velocity (jg), liquid superficial velocity (jf), and void fraction (α). The void fraction is an important factor in IAC prediction, and logarithmic transformation preprocessing of training data can significantly improve the model's predictive ability for real data.ConclusionsThe results of this study provide reference for future interpretability research on interface area concentration.

The two-phase flow is a vital topic of thermal-hydraulic research in light-water nuclear reactors. The fluid dynamics and heat transfer characteristics of two-phase flow impact the safety and efficiency of nuclear power plants. The two-phase flow results in various transients in the nuclear reactor: The change of the pressure drop redistributes the flow rate of the channels, and the generation of the void in the core alters the reactivity and consequently the power. The two-phase flow is among the most complex flows since the interface structure evolves and the mass/momentum/energy transfers across the interface. The two-phase flow research of the nuclear system started in the 30s and 40s of the last century, when the pioneer scientists studied the flow instability and the pressure drop. Along with the rise of commercial nuclear reactors in the 50s, the interests lay in the critical flow, the critical heat flux, and the void distribution. In the 70s, the industry shifted its interest to the blow and the flooding, which are compound flow phenomena in specific geometries of the reactor system. In the 90s, the focus was to improve the ability of the nuclear safety codes, such as developing the constitutive models and conducting benchmark experiments. Since this century, the growth of the measuring techniques has supported fundamental research, like the interface evolution and the behavior under extreme conditions. Also, the development of the computers enables the multi-dimensional study using computational fluid dynamics codes and enables the multi-physics study using the codes-coupling platform. This paper aims to understand the research trend of the two-phase flow in the nuclear system by reviewing the history, summarizing the critical problems, and presenting the status. Five phenomena, which lack knowledge and need to be investigated, are described in the current review: the two-phase flow structure, the interfacial drag, the heat transfer crisis, the interphase heat/mass transfer, and the counter-current flow limitation. Three research directions are suggested: advanced modeling, advanced measuring techniques, and cross-cutting fields. Concluding remarks, with research recommendations from the local scale to the system scale, are given to the two-phase flow in the nuclear system.

Core make-up tank is a vital component of the reactor safety system. Under accident conditions, dramatic direct contact condensation may occur in the core make-up tank, causing pressure decrease or oscillation, which would be harmful to its safety injection function. To enhance the prediction accuracy of the process, it is necessary to build a special model to calculate the direct contact condensation phenomenon in the core make-up tank, since in the core make-up tank the steam would go through a steam space before it reaches the liquid surface, which is quite different from the mostly studied conditions, where a pipe is often used to guide the steam to the water. In this paper, the jet flow velocity distribution theory was first employed to estimate parameters like the steam velocity profile and the maximum steam velocity at the interface, and then the hypothesis nozzle analysis method was utilized to get the hypothesis nozzle diameter, the effective velocity, the steam penetration length, the effective steam condensation fraction of the jet flow region, etc. Besides, to properly evaluate the heat transfer temperature difference, based on the conversation of energy, a multi-layer model was used to take the temperature stratification effects into account, and the whole liquid region was further divided into two layers, the hot layer and the cold layer. In this way the heat transfer temperature difference was corrected. Considering the effective steam condensation fraction of the jet flow region and natural circulation region near the water surface, a specific condensation heat transfer model was established based on the interface transport model for the jet flow region and the McAdams correlation for natural circulation region. With the help of the effective steam condensation fraction of the jet flow region, the new model could be used to predict the interface heat transfer of liquid and steam during the whole process of core make-up tank safety injection, including the complete condensation period, the partial condensation period and the natural circulation period. Several sets of transient data from two different experiment systems were used to test the new model, with the core make-up tank initial temperature ranges from 21 to 61 ℃, and steam generator pressure from 0.15 to 8.7 MPa. It is found that the experimental data could be well predicted, and a clear improvement compared to the origin model of Relap5 can be observed, which preliminarily demonstrates the validation of the model. This research may be helpful for the accurate prediction of core make-up tank safety injection process and the safety analysis of related accidents.

The guided waves in the elbow was analyzed using semi- analytical finite element method, which was used to calculate the dispersion curves, along with the vibration deformation and energy flow density distribution of the elbow cross-section. The L(0,1) mode guided wave was used to detect different defects in the U-shaped pipeline in experiment. The average experimental velocity of L(0,1) mode in U-shaped pipeline was between the theoretical velocities of the straight pipe and the elbow, due to the different velocities in the two parts. The detection sensitivity difference of L(0,1) mode guided wave to defects in intrados and extrados of the elbow was analyzed, the experimental result showed that the reflected signal amplitude of defect in extrados was significantly larger than the defect in intrados, which was consistent with the energy flow density distribution in elbow cross-section.