• Water Engineering Research
• /
• v.6 no.1
• /
• pp.1-15
• /
• 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.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.30 no.10 s.253
• /
• pp.942-949
• /
• 2006

A lumped model is proposed to predict liquid ejection characteristics of a thermally driven inkjet print head. The model is based on a two-dimensional heat conduction equation, an empirical pressure-temperature equation and a nonlinear hydraulic flow-pressure equation. It has been simulated through the construction of an equivalent R-C circuit, and subsequently analyzed using SIMULINK and a circuit simulation tool, PLECS. Using the model, heating and cooling characteristics of the head are predicted to be in agreement with the IR temperature measurements. The effects of the head geometry on the drop ejection are also analyzed using the nonlinear hydraulic model. The present model can be used as a design tool for a better design of thermal inkjet print heads.

• Journal of Energy Engineering
• /
• v.10 no.3
• /
• pp.294-298
• /
• 2001

A research for transmutation reactor is in progress to transmute high radioactive isotopes into low radioactive ones. In this study, thermal-hydraulic and structural analysis was performed to design liquid metal target system that would be used in subcritical transmutation reactor. Diffuse plate installation was considered to enhance cooling of window. And thermal-structural analysis of window was performed varying window thickness, beam power, and coolant flow rate to determine target system design valuers. It is ensured that maximum window temperature and stress would be acceptable in the design condition.

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

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1504-1513
• /
• 2019

Recently, ocean nuclear reactors have received attention due to enhanced safety features. The movable and transportable characteristics distinguish ocean nuclear reactors from land-based nuclear reactors. Therefore, for safety/design analysis of the ocean reactor, the thermos-hydraulics must be investigated in the moving system. However, there are no studies reporting the general two-fluid equations that can be used for multi-dimensional simulations of two-phase flows in moving systems. This study is to systematically formulate the multi-dimensional two-fluid equations in the non-inertial frame of reference. To demonstrate the applicability of the formulated equations, we perform a total of six different simulations in 2D tanks with translational and/or rotational motions.

• Nuclear Engineering and Technology
• /
• v.55 no.9
• /
• pp.3320-3325
• /
• 2023

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re = 2,409 and Re = 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

• Economic and Environmental Geology
• /
• v.39 no.5 s.180
• /
• pp.537-554
• /
• 2006

Water levels of thermal groundwater ($>30^{\circ}C$) were recorded from March 2002 to June 2006 at several monitoring wells within the Yuseong spa area. Using these data, we elucidated the long-term cyclic fluctuations of thermal groundwater levels with 1 year period. We also observed a noticeable water level variation with periods of 0.5, 1 and 7 days in most monitoring wells, which indicates relatively good hydraulic connectivity within the main hotspring area. By comparing water level variations among several wells, we found out that E-W and N-S trending geological structures should be an important control factor for emplacement and flow of thermal groundwater in the study area. It may be also inferred that geothermal source is highly associated with the hydraulic connectivity of aquifers at the Yuseong spa area.

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

• Nuclear Engineering and Technology
• /
• v.49 no.5
• /
• pp.968-978
• /
• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• Nuclear Engineering and Technology
• /
• v.41 no.5
• /
• pp.655-676
• /
• 2009

The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.