Understanding the behavior of the primary coolant in a nuclear reactor is crucial for evaluating thermal margins and ensuring reactor safety. This study examines the behavior of the primary coolant using the ACOP experimental facility, a 1/5 scale model of the APR1000 reactor, to evaluate the effects of flow rate and flow unbalance on coolant distribution. Experiments were performed at flow rates of 60 %, 80 %, 100 %, and 110 % under balanced conditions, as well as under 15 % unbalanced and extreme unbalanced conditions with three pumps. The coolant flow characteristics were assessed by measuring the core inlet flow distribution, core outlet pressure, upper plenum pressure, and hot leg velocity using 177 core simulators. The results showed that the core inlet flow distribution ranged between 88 % and 112 % across all conditions. Under extreme unbalanced conditions, a maximum deviation of 5.3 % in single core flow distribution was observed. The outlet pressure distribution remained stable under four-pump conditions but showed greater variation with three pumps. The study concludes that coolant flow rate and flow unbalance had minimal impact under balanced four-pump conditions, while significant effects on outlet pressure, upper plenum pressure, and hot leg velocity were evident under extreme unbalanced conditions.

Small leaks within a nuclear power plant can escalate into significant leaks, leading to plant shutdown and substantial losses if operational limits are exceeded. Thus, the demand for systems that can rapidly detect minor leaks is increasing. Current research seeks to address this need. Understanding the thermal–hydraulic features of these systems is crucial for their evolution. In this study, we designed a numerical analysis process to evaluate a prototype leak detection system simulator. We devised a numerical model to simulate leaks and subsequently conducted a bifurcated computational fluid dynamics analysis to assess its viability. First, we examined the leak dynamics within the insulation, focusing on the effects of the gap between the pipe’s insulation and its external casing. The findings from the initial analysis informed the conditions for studying the collection loop. The assessment highlighted a shift in relative humidity downstream of the collection loop, mirroring experimental observations. The results suggest that the newly developed small leak detection system can effectively detect leaks by collecting the escaping fluid through a collection loop and analyzing variations in relative humidity when a small leak occurs in the high-pressure piping of a nuclear power plant. These findings have been instrumental in developing a collection loop experimental apparatus, which will further elucidate the heat transfer dynamics during its operation.

The flow distribution characteristic at the reactor core is essential for the licensing and design process. An hydraulic test facility was established to simulate the internal flow of the APR1000 reactor at a reduced scale of 1/5. The design concept of the test facility with an applied scaling method was introduced briefly, and the effect of the lower support structure on the core inlet flow distribution was investigated experimentally, as a preliminary test. The flow characteristics were quantified and discussed from the directly measured flow rates of 177 fuel assemblies.

The experimental data of core flow distribution are indispensable for obtaining licensing and facilitating the design of fluid systems of nuclear reactors. In this study, an Advanced power reactor Core flow and Pressure (ACOP) test facility was established to experimentally simulate the internal flow of the Advanced Power Reactor 1000 (APR1000) on a reduced length scale of 1/5. The core region was simulated by using 177 core simulators representing the fuel assemblies of the APR1000. The APR1000 flow distributions were synthetically identified by accurately measured parameters: the core inlet flow rate and outlet pressure under the four-pump balanced and unbalanced flow conditions. The overall inlet flow rates ranged from 87.7% to 112.0% relative to the averaged flow rate. Here, we scrutinize the flow distributions considering the flow conditions and internal structures and briefly describe the applied scaling method and design concept of the test facility.

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.

Suppose a small amount of leakage exceeds the limit of the operating conditions of a nuclear power plant. In that case, losses due to the nuclear power plant shutdown may occur, or accidents due to leakage may occur. Consequently, there has been a need to develop a system capable of quickly detecting even small amounts of leakage, and research has been conducted to address this need. To develop such a system, it is necessary to understand the thermal-hydraulic characteristics of the system. This study established a CFD-based evaluation process for evaluating leak detection systems. A numerical analysis model was used to simulate leakage, and a two-step CFD analysis was performed to assess its applicability. In the first step, a study of the leakage behavior in the insulation material was conducted, and the effect of the gap between the pipe insulation material and the outer cover was evaluated. In the second step, the collection loop was analyzed using the results of the leakage behavior analysis in the insulation as an input condition. The results of this study were used to construct the collection loop experimental device and will help in understanding the heat flow characteristics in the collection loop.

Flow mixing between adjacent subchannels within a wire-wrapped hexagonal fuel rod bundle affects radial heat transfer and determines the maximum cladding temperature, which is a key parameter to ensure the fuel safety margin in a sodium-cooled fast reactor (SFR). In the ENERGY model, both the effec-tive eddy diffusivity and the edge swirl velocity ratio are employed to characterize the subchannel flow mixing. The Korea Atomic Energy Research Institute conducted subchannel flow mixing tests on the 37 -pin and 61-pin bundle assemblies. The test assemblies were fabricated by reflecting thermal-hydraulic similarities of the Prototype Gen-IV SFR candidate cores. The subchannel flow mixing experiment was conducted using a wire-mesh sensing system. The subchannel flow mixing coefficients are determined by the least-square method so that the difference between the SLTHEN code prediction and experimental data is minimized numerically. The results show good agreement with the previous correlations, espe-cially with the Cheng-Todreas model. (c) 2021 The Author(s). Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).

A meta-stable effect on a subcooled critical flow in orifice was investigated. For an investigation of the meta-stable effect, two homogeneous nucleation models, e.g., the Alamgir and Lienhard’s model and the Elias and Chambre’s model, and one isentropic model, e.g., the Henry and Fauske’s model, were compared with respect to previous test data performed by Sozzi and Sutherland. From comparisons of throttle pressures, the Alamgir and Lienhard’s model provided a lower bound and the Elias and Chambre’s and the Henry and Fauske’s, an upper bound. The experimental data were lying between the two bounds. The well-known diameter effect in critical flow was investigated and rather, diameter ratio (d/D) effect was found important in critical flow.

Because steam leakage accidents at nuclear power plants could cause various problems such as radiation leakage into the environment and nuclear plants shutdown, it is very important to monitor steam leakage from the safety point of view. The steam leakage sampling system consists of the collection, transfer, and analysis processes for the humid air. The collected humid air was diffused by diffusion and convection in the collection and transport process in the sampling system. Therefore, it is necessary to understand the flow characteristics of the humid air in the sampling system for measuring and analyzing the collected humid air accurately. In this study, we established a numerical methodology for simulating the mixing, diffusion of gas mixture during collection and transportation process. Then diffusion characteristics of humid air were predicted using numerical analysis based on the established methodology, also the behavior of humid air during the transportation process was predicted.

Because steam leakage in nuclear power plants could cause various problems including nuclear plants shutdown and leakage of radioactive materials, it is very critical to monitor steam leakage from the safety point of view. The steam leakage collection system consists of the collection, transfer, and analysis processes for the humid air. The relative humidity of the collected humid air could be reduced by the diffusion and convection during the collection and transfer process. Therefore, it is necessary to understand the diffusion and convection characteristics of the humid air for measuring the relative humidity of the collected humid air accurately. In this study, we established a numerical methodology for simulating the mixing, diffusion and turbulence of gas mixture. Then diffusion and convection characteristics of the humid air were predicted using numerical analysis based on the established methodology.

The separation efficiency of the moisture separator in a steam generator is the most important key parameter to ensure the safety of turbine and to obtain high efficiency in a nuclear power plant. The objective of this research is to investigate which experimental method can be used for evaluating the performance of the separator, namely moisture carry over (MCO), of the prototype separator. The air/water test facility based on the similarity law was constructed and developed an experimental method to quantify the MCO. For a various mass quality ranged from 0.277 to 0.382, several experiments were carried out to evaluate the MCO and to verify the experimental method. The obtained results were compared to data on the prototype condition and showed a good agreement with that of the prototype conditions.

A moisture separator is an essential component in PWR steam generator. The performance of separator is the most important to assure acceptable steam quality for the operation efficiency in nuclear power plant. The major parameter related to the performance of moisture separation system (MSS) are the moisture carryover (MCO) rate and pressure loss through the MSS. KAERI (Korea Atomic Energy Research Institute) has developed the experimental method to evaluate the performance of Westinghouse (WH) type separator using a scaling law with the reduced models in low pressure air-water test facility. However, a considerable investigation is required to simulate the complex fluid flow of a mixture of two-phase. Besides, there is no verified experimental method based on a scaling law, even though a large number of studies have been carried out. As a preliminary study, in this paper, a numerical analysis is carried out to validate the similarity for the internal flow under a single-phase flow, since the flow field can be assumed as a stokes flow if the only small droplets have low relative velocity to the gas flow, and the single-phase flow field in the internal region is of great importance to better understand the momentum and steam-water separating mechanism.

In a multidimensional two-phase flow analysis, bubble size significantly affects interfacial transfer terms such as mass, momentum, and energy. With regard to bubbly flow, the application of a simple correlation-type bubble size model presents certain advantages, including short calculation times and ease of usage. In this study, we propose a semi-theoretical correlation developed from a steady state bubble number density transport equation for predicting the distribution of local bubble size using a computational fluid dynamics (CFD) code. The coefficients of the new correlation were determined using the local bubble parameters obtained on the basis of three existing vertical air-water experiments. Finally, these were implemented in commercial CFD code and evaluated against experimental data, which showed that the proposed correlation exhibits good prediction capability for forced convective air-water bubbly flows under low pressure conditions.

A test facility to investigate the flow characteristics inside reactor vessel for the Prototype Gen-IV Sodium-cooled Fast Reactor was constructed. In the test facility, reactor vessel and the main in-vessel components are linearly reduced at a scaling ratio of 1/5 and water is used as the working fluid. In the reactor vessel of the test facility, the main components such as core, UIS, PHTS pump and IHX are installed. The exteriors of the main components are conserved following the scaling ratio of 1/5, but the internal flow paths inside the fuel assemblies and IHXs are uniquely designed for precise measurement of flow rate and conserving the pressure drop characteristics. In order to determine the configuration and specific dimension of the internal flow path, separate analysis and experiment for validation are required. In the present paper, the basic design concept of internal flow path for fuel assembly simulator used for the reactor flow distribution test was established, and the detailed size of main design factors were estimated using commercial CFD code. The internal flow path of the simulator is composed of a receptacle, variable-resistance rotating orifice spool, venturi tube, and connection lines. The flow rate through the simulator could be estimated by measuring the differential pressure between inlet section and the throat of the venturi tube. The orifice spool is installed at the downstream of the receptacle, by which entire pressure drop through the simulator is controlled. A series of CFD analysis was conducted to estimate the throat diameter of the venturi tube and the size of the holes at the orifices. The geometry of the orifice is determined to obtain target pressure drop when the angle between two orifice plates is from 0 degrees to 45 degrees. The design specifications were applied to the fabrication of the fuel assembly simulators, and the performance of them was verified experimentally. The pressure drop of 112 fuel assembly simulators was successfully adjusted to be within +/- 1% of the target pressure drop. The relationship between the mass flow rate and differential pressure of the venturi tube was also obtained, and empirical correlation was suggested.

The flow characteristics at the shell side of a prototype intermediate heat exchanger (p-IHX) in a prototype generation-IV sodium-cooled fast reactor (PGSFR) were investigated experimentally in this study. The shell side of the p-IHX consisted of an inlet window, tube bundles, grid plates, and an exit flow channel. Two different test facilities were newly constructed to identify the pressure drop characteristics due to the complicated configuration with minimized scaling distortion. An intermediate heat exchanger test loop for PGSFR (iHELP) was constructed to characterize the flow resistance of the tube bundle regions with the grid plates and inlet window. The iHELP test section with a slab-shaped cross section was manufactured with a 1/29.6 vol scale ratio by preserving the height of the p-IHX. The hydraulic diameter at the tube bundle regions, the porosity of the grid plates, and the configuration of the tube bundles were conserved. In addition, the Reynolds number was preserved, which determined the flow rate conditions for the iHELP experiments. To investigate the pressure drop characteristics of the exit flow channel (EC), an intermediate heat exchanger test loop for the exit flow channel (IEC) was built with a 1/5 linearly reduced length scale by preserving the geometry of the flow path based on the EC of the p-IHX. The Euler number was conserved with a 1/8 Reynolds number ratio corresponding to a 1/1 flow velocity ratio to the p-IHX for dynamic similarity. By using each test facility the experimental pressure drop data satisfying the experimental uncertainty requirements were obtained separately for the bundle and exit channel regions. The pressure drop at the shell side of the p-IHX under a wide range of flow rate conditions was determined by using the pressure drop results from the iHELP and the IEC test facilities, which would be useful in improving and validating the pressure drop correlations in the IHX design computational code (SHXSA).

To predict leakage flowrate from major equipment/component in reactor system, potential leakage points in reactor system were reviewed and models of critical flowrate were investigated. For predictions of leakage flow, three kinds of critical flow models, e.g., ideal gas flow model, Trapp-Ransom model, and Henry-Fauske model, were investigated. Using Henry-Fauske model, leakage flows were obtained for selected leakage points in reactor system and equivalent crack sizes were also estimated for 0.5 gpm leakage.

A CFD analysis was performed for the full scale model and a 1/5-linear scale model to evaluate the similarity of boric acid mixing in the downcomer and the reactor core. By applying the linear scale methodology and conservation of Euler number, the prototype flow distribution is preserved. Since the core of a prototype nuclear power plant is composed of a mixing grid of nuclear fuel assemblies with complex shapes, a large-capacity computer is required to simulate the actual shape by CFD. In this study, the core of the reference nuclear reactor was modeled as a porous media and preserved axial pressure drop and transverse flow distribution characteristics. The similarity of the 1/5-linear scale model was compared with the flow distribution of the reference full scale plant. The 1/5-linear scale core was evaluated for two models: (1) Free Cavity and (2) mixing vane. The free cavity core model is a model in which the active core part is completely empty. The mixing vane core model is a model in which the pressure drop of the mixing vane in the fuel assembly (F/A) and the hydraulic diameter of the fuel rod are preserved. That is, the F/A shape of the core is simulated closely and pressure drop and cross flow characteristics are also preserved. The similarity of boric acid to Shutdown Cooling System (SCS) and Control Volume-and-Chemical System (CVCS) was well evaluated.