• 터널과지하공간
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• 제8권1호
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• pp.17-25
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• 1998

This paper investigates coupled thermal, mechanical and hydraulic phenomena in deep rock mass especially for underground heat storage system. Firstly, concepts of underground heat storage were presented and coupling phenomena in this area were illustrated. In order to understand the basic mechanism of thermal, hydraulic and deformation behavior in rock cavern disturbed by thermal gradient about 10$0^{\circ}C$, various numerical experiments were conducted using several codes. The study involves the behavior of fractured rock mass including rock joint. In spite of the limitation of codes modelling fully coupled effects, these codes could be applied in analysis of underground heat storage. The heat loss in rock mass, which is a major factor in heat storage, is insignificant in all results.

• Water Engineering Research
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• 제6권1호
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• pp.1-15
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• 2005

The numerical model, FLUENT, was employed to investigate the effect of the heated water discharged from the diffuser of Boryung Power Plant. Temperature patterns of the thermal effluent discharged from two proposed types of the diffusers was evaluated for maximum flood and maximum ebb tide. The hydraulic model experiments were also performed in the reduced scale of 1/150 to verify the numerical simulation results. The buoyant jets discharged from the diffusers were found to be significantly affected by the ambient flows beyond the region where the effluent momentum was dissipated. Both the numerical and experimental results showed that the area of the excess isotherm for Type 1 diffuser was larger than that for Type 2 diffuser. Type 2 diffuser system was observed to be a more effective diffuser design than Type 1 diffuser system based on the temperature reduction and excess isotherm obtained from the numerical simulation in the ambient flows.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1263-1274
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• 2009

Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3182-3195
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• 2021

The implementation and validation of multi-dimensional (multi-D) features in thermal-hydraulic system codes aims to extend the application of these codes towards multi-scale simulations. The main goal is the simulation of large-scale three-dimensional effects inside large volumes such as piping or vessel. This novel approach becomes especially relevant during the simulation of accidents with strongly asymmetric flow conditions entailing density gradients. Under such conditions, coolant mixing is a key phenomenon on the eventual variation of the coolant temperature and/or boron concentration at the core inlet and on the extent of a local re-criticality based on the reactivity feedback effects. This approach presents several advantages compared to CFD calculations, mainly concerning the model size and computational efforts. However, the range of applicability and accuracy of the newly implemented physical models at this point is still limited and needs to be further extended. This paper aims at contributing to the validation of the multi-D features of the system code ATHLET based on the simulation of the Tests 1.1 and 2.1, conducted at the test facility ROCOM. Overall, the multi-D features of ATHLET predict reasonably well the evolution from both experiments, despite an observed overprediction of coolant mixing at the vessel during both experiments.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.11-32
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• 2006

System thermal-hydraulic codes have been used in the past decades in the areas of design, operation, licensing and safety of Nuclear Power Plants (NPPs). The development and validation of these codes have reached a high degree of maturity, through the consideration of huge experiments and advanced numerical models. Nowadays, the analyses are based upon realistic approaches rather than the conservative evaluation models. However the applications of these computational tools require preliminary qualification issues. Although huge amounts of financial and human resources have been invested for the development and improvement of codes, the calculation results are still affected by errors. In the sophisticated nuclear technology, design and safety of NPP, these errors must be quantified. An overview of the state of the art of the current thermal-hydraulic system code is developed and the need of uncertainty analysis in code calculations is emphasized. Several sources of uncertainty have been classified and commented, and typical applications of such methods are shown.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3874-3897
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• 2023

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High-Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next-generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds-averaged Navier-Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.1126-1135
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• 2022

The MARS code has been assessed for the prediction of onset of flow instability (OFI) in a vertical channel. For assessment, we built an experiment database that consists of experiments under various geometry and thermal-hydraulic condition. It covers pressure from 0.12 to 1.73 MPa; heat flux from 0.67 to 3.48 MW/m²; inlet sub-cooling from 39 to 166 ℃; hydraulic diameters between 2.37 and 6.45 mm of rectangular channels and pipes. It was shown that the MARS code can predict the OFI mass flux for pipes reasonably well. However, it could not predict the OFI in a rectangular channel well with a mean absolute percentage error of 8.77%. In the cases of rectangular channels, the error tends to depend on the hydraulic diameter. Because the OFI is directly related to the subcooled boiling in a flow channel, we suggest a modified subcooled boiling model for better prediction of OFI in a rectangular channel; the net vapor generation (NVG) model and the modified wall evaporation model were modified so that the effect of hydraulic diameter and heat flux can be accurately considered. The assessment of the modified model shows the prediction of OFI mass flux for rectangular channels is greatly improved.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.54-67
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• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1125-1134
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• 2024

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

• 대한기계학회논문집B
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• 제33권9호
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• pp.643-657
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• 2009

This paper introduces the current nuclear experimental research activities in KAERI, the unique nuclear research institute in Korea, and the universities in Korea to solve and assess the issues which have been faced in the design of new reactors such as APR1400, SMART, GEN-IV reactors as well as fusion reactor. Also the experimental evaluations of safety for operating nuclear plants have been presented. The nuclear thermalhydraulic experiments performed in such organizations are classified the fundamental test, the separated effect test, and the integral effect test with ATLAS and SNUF. Introduction is deployed according to institutes. Finally, the future works and the direction of research voyage in the nuclear thermal-hydraulic field were suggested.