This study develops and validates a Reduced Order Model (ROM) integrated with Digital Twin technology for real-time temperature control in the Main Control Room (MCR) of a nuclear power plant. Utilizing Computational Fluid Dynamics (CFD) simulations, we obtained detailed three-dimensional thermal flow distributions under various operating conditions. A ROM was generated using machine learning techniques based on 94 CFD cases, achieving a mean temperature error of 0.35%. The ROM was further validated against two excluded CFD cases, demonstrating high correlation coefficients (R > 0.84) and low error metrics, confirming its accuracy and reliability. Integrating the ROM with the Heating, Ventilating, and Air Conditioning (HVAC) system, we conducted a two-month simulation, showing effective maintenance of MCR temperature within predefined criteria through adaptive HVAC control. This integration significantly enhances operational efficiency and safety by enabling real-time monitoring and control while reducing computational costs and time associated with full-scale CFD analyses. Despite promising results, the study acknowledges limitations related to ROM’s dependency on training data quality and the need for more comprehensive validation under diverse and unforeseen conditions. Future research will focus on expanding the ROM’s applicability, incorporating advanced machine learning methods, and conducting pilot tests in actual nuclear plant environments to further optimize the Digital Twin-based control system.

Mechanical seals, such as labyrinth seals, are typically installed at the turbine outlets to prevent oil leakage. However, these seals undergo deformation because of the vibrations of the rotor, even during normal turbine operating conditions, which may cause an increase in oil leakage. In this study, the oil leakage performance of three labyrinth seals with different types of seal teeth, narrow stainless teeth (Type 1), wide aluminum teeth fixed on the body (Type 2), and fixed wide aluminum movable teeth (Type 3), were evaluated using finite element (FE) and computational fluid dynamics (CFD) analyses. Three-dimensional FE models of the rotor and oil deflectors were developed, and the plastic deformation of the teeth of the labyrinth seals was predicted when the rotor impacted the sealing teeth during turbine operation. The oil leakage was predicted using CFD analysis. The results indicated that the Type 3 seal, including movable teeth, is beneficial in preventing leakage and tooth deformation compared with the other types. The Type 2 seal is advantageous because it results in a smaller increase in gap size and greater vena contracta effects than the Type 1 seal. The results of this study could be helpful when designing and selecting the teeth of a labyrinth seal.

The core inlet flow distribution in the APR1000 reactor is critical for ensuring the reactors safety and efficient operation by maintaining uniform coolant flow across fuel assemblies. Previous studies, though insightful, faced challenges in fully replicating reactor-scale flow conditions due to technical and economic constraints associated with scaled-down experimental models and the limited numerical validation methodologies. This study addresses these limitations by developing and validating a robust computational fluid dynamics (CFD) methodology to accurately analyze the core inlet flow distribution. A 1/5 scaled-down experimental model adhering to similarity laws was employed for validation. CFD analyses using ANSYS Fluent and CFX, combined with turbulence model evaluations and grid sensitivity studies, demonstrated that the SST and RNG k-ε turbulence models provided the most accurate predictions, with a high correlation to previous experimental data. Full-scale simulations revealed uniform coolant distribution at the core inlet, with peripheral assemblies exhibiting higher flow rates, consistent with previous experimental observations. Quantitative metrics such as the coefficient of variation (COV), relative error (RD), and root mean square error (RMSE) confirmed the superior performance of the SST model in CFX, achieving a COV of 7.993% (experimental COV: 5.694%) and an RD of 0.047. This methodology not only validates the CFD approach but also highlights its applicability to reactor design optimization and safety assessment. The findings of this study provide critical guidelines for analyzing complex thermal-fluid systems in nuclear reactor designs.

The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics (CFD) analysis methodology was incorporated as a first step to establish an RCS natural circulation evaluation technique to generate RCS natural circulation input parameters for the MELCOR analysis of thermally induced steam generator tube rupture (TI-SGTR) in nuclear power plants. Benchmarking tests were conducted against existing experimental studies; the results demonstrated a difference of 9.4% or less between the experimental and CFD analysis results with respect to the main evaluation factors. Subsequently, a steam generator tube simplification modeling technique was established for application to nuclear power plants, and CFD analysis was conducted to determine its applicability. The CFD analysis results revealed that when numerous tubes are simplified into one equivalent tube, the thermal flow characteristics generated in the RCS could be distorted. The findings of this research are expected to be helpful in understanding the thermal flow characteristics of natural circulation in the RCS. Further, the findings may potentially serve as a foundation for future CFD analysis research related to the natural circulation in the RCS of nuclear power plants.

In the present study, a series of numerical analysis was performed to identify the flow characteristics inside the CE-type moisture separator applied to domestic steam generators. To investigate the similarity between normal operation condition of the prototype separator and experimental condition using air-water test facility with real scale and 1/2 scale test sections, single phase and two-phase calculations were carried out. The results of the single phase calculations showed that the outward flows through the flow holes on the wall turned out to be similar trend regardless of the working fluid and inlet conditions on real scale geometry. However, the pressure loss coefficient for the 1/2 scale geometry was relatively larger than that for the real scale geometry. In the two-phase calculation, Eulerian-Lagrangian method was adapted to investigate the behavior of the droplets. Two extreme conditions for the post impingement behavior of the droplet, which were escape and rebound conditions; were applied and the results between them were compared.

Owing to pipe thinning, fatigue damage, and aging, pipes, valves, and devices installed in the primary and secondary systems of nuclear power plants may leak high-temperature/high-pressure reactor coolant. Thus, a system must be developed to determine if the leakage is exceeding the operating limit of the nuclear power plant, thereby mitigating any loss of life or economic loss in such cases. In this study, a validated numerical analysis method was established to initially simulate the leakage behavior and subsequently to evaluate the small amount of leakage in the compartment. For this purpose, a vapor-jet collision test in the compartment and a vapor-jet test in the pipe were performed; numerical analysis was conducted, and comparative analysis was performed to verify the validity of the established method. The evaluation results suggested that the proposed numerical analysis method could optimally simulate the flow characteristics of the steam jet. Notably, compared to the existing evaluation method, the proposed approach simulated a more detailed behavior of the jet formed at the leakage point. In future research, the results of this study (data) will be used to inform the design of the second phase of the leak-capture system and will be served as the foundation for a performance-optimization study on the capture system.