• Water Engineering Research
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• v.6 no.1
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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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• v.55 no.4
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• pp.1555-1562
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• 2023

Natural circulation phenomena have been nowadays largely revisited aiming to investigate the performances of passive safety systems in carrying-out heat removal under accidental conditions. For this purpose, assessment studies using CFD (Computational Fluid Dynamics) and also 3D thermal-hydraulic system codes are considered at different levels of the design and safety demonstration issues. However, these tools have not being extensively validated for specific natural circulation flow regimes involving flow mixing, temperature stratification, flow recirculation and instabilities. In the present study, an experimental test case based on a small-scale pool test rig experiment performed by Korea Atomic Energy Research Institute, is considered for code-to-code and code-to-experimental data comparison. The test simulation is carried out using the FLUENT and the 3D thermal-hydraulic system CATHARE-2 codes. The objective is to evaluate and compare their prediction capabilities with respect to the test conditions of the experiment. It was observed that, notwithstanding their numerical and modelling differences, similar agreement results are obtained. Nevertheless, additional investigations efforts are still needed for a better representation of the considered phenomena.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.119-128
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• 2006

We start to develop a predictable technology for thermal-hydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing the advanced interface tracking method to improve the conservation of volume of fluid. The present paper describes a part of the development of the twophase flow simulation code TPFIT with the advanced interface tracking method. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results of numerical simulation, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values obtained by the advanced neutron radiography technique including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.

• Journal of the Korean Society for Geothermal and Hydrothermal Energy
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• v.1 no.1
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• pp.1-8
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• 2005

Aquifer Thermal Energy Storage (ATES) can be a cost-effective and renewable geothermal energy source, depending on site-specific and thermohydraulic conditions. To design an effective ATES system having the effect of groundwater movement, understanding of thermohydraulic processes is necessary. The heat transfer phenomena for an aquifer heat storage are simulated by using FEFLOW with the scenario of heat pump operation with pumping and waste water reinjection in a two layered confined aquifer model. Temperature distribution of the aquifer model is generated, and hydraulic heads and temperature variations are monitored at the both wells during 365 days. The average groundwater velocities are determined with two hydraulic gradient sets according to boundary conditions, and the effect of groundwater flow are shown at the generated thermal distributions of three different depth slices. The generated temperature contour lines at the hydraulic gradient of 0.001 are shaped circular, and the center is moved less than 5 m to the direction of groundwater flow in 365 days simulation period. However at the hydraulic gradient of 0.01, the contour center of the temperature are moved to the end of east boundary at each slice and the largest movement is at bottom slice. By the analysis of thermal interference data between two wells the efficiency of the heat pump system model is validated, and the variation of heads is monitored at injection, pumping and no operation mode.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.213-218
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• 1996

A simulation model for the transient behavior of CANDU U-tube steam generator(UTSG) has been developed for application to the simulation of operational transient behavior of CANDU nuclear power plant. For application to CANDU UTSG. tile design characteristics of CANDU UTSG such as Wolsong Units, feedwater inlet near the tube sheet. is approximated. For realistic prediction of thermal hydraulic behavior of and tube bundle region is divided into two separate control volumes, subcooled region and saturated region. and the variation of thermal hydraulic properties within a control volume is considered. Test results for typical CANDU operational transient case show reasonable transient behavior of steam generator and considered to be applicable to the simulation of overall plant.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.561-586
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• 2012

This paper describes a summary of final prediction results by system-scale safety analysis codes during the OECD/NEA/CSNI ISP-50 exercise, targeting a 50% Direct Vessel Injection (DVI) line break integral effect test performed with the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). This ISP-50 exercise has been performed in two consecutive phases: "blind" and "open" phases. Quantitative comparisons were performed using the Fast Fourier Transform Based Method (FFTBM) to compare the overall accuracy of the collected calculations. Great user effects resulting from the combination of the possible reasons were found in the blind phase, confirming that user effect is still one of the major issues in connection with the system thermal-hydraulic code application. Open calculations showed better prediction accuracy than the blind calculations in terms of average amplitude (AA) value. A total of nineteen organizations from eleven countries participated in this ISP-50 program and eight leading thermal-hydraulic system analysis codes were used: APROS, ATHLET, CATHARE, KORSAR, MARS-KS, RELAP5/MOD3.3, TECH-M-97, and TRACE.

• Journal of the Korean Society for Geothermal and Hydrothermal Energy
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• v.13 no.1
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• pp.14-20
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• 2017

Groundwater heat pump (GWHP) system can achieve higher performance of the system by utilizing heat source of the annual constant groundwater temperature. The performance of GWHP system depends on the ground thermal environment such as groundwater temperature, groundwater flow rate and hydraulic conductivity. In this study, the geothermal environment was analyzed by using numerical simulation for develop the two-well geothermal system. As the result, this paper shows the change of the groundwater level and underground temperature around wells according to the conditions of flow rate and hydraulic conductivity.

• 한국전산유체공학회:학술대회논문집
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• 2003.10a
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• pp.205-206
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• 2003

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.426-441
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• 2003

A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

• Journal of Energy Engineering
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• v.13 no.1
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• pp.51-59
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• 2004

The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited compulsational capability at that time, they usually used very simplified physical models for a real-time simulation of NSSS thermal-hydraulic transients, which entails inaccurate results and, thus, the possibility of so-called "negative training", especially for complicated two-phase flows in the reactor coolant system. In resolve the problem, KEPRI developed a realistic NSSS T/H program ARTS which was based on the RETRAN-3D code for the improvement of the Nuclear Power Plant full-scope simulator. The ARTS (based on the RETRAN-3D code) guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there is some possibility of failing to calculate in the case of large break loss of coolant accident (LBLOCA) and low-pressure low-flow transient. In this case, the backup calculation system cover automatically the ARTS. The backup calculation system was expected to provide substantially more accurate predictions in the analysis of the system transients involving LBLOCA. The results were reasonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with FSAR and the AMSI/ANS-3.5-1998 simulator software performance criteria.