• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.18 no.5
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• pp.402-409
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• 2006

For the design of a liquid direct contact cooling system, thermal and hydraulic analysis has been carried out. Well-known Zukauskas correlations are used to estimate the Nusselt number between the liquid refrigerant columns and the inlet airflow. The inlet air velocity is set at a typical value used in an actual showcase. For a constant column number, the best nozzle arrangement is determined for the maximum heat transfer. Heat transfer increases as the transverse pitch of the refrigerant column decreases. Among all the cases dealt with in the present study, the staggered arrangement with 140-columns of $14{\times}10$ shows the best thermal peformance and the expected temperature drop is $27.8^{\circ}C$. The effect of downstream refrigerant columns on the overall thermal performance is investigated as well.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• Journal of the Korean Society for Geothermal and Hydrothermal Energy
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• v.11 no.4
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• pp.28-34
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• 2015

The groundwater heat pump (GWHP) systems have great potential for heating-cooling system which use annual constant groundwater temperature for heat source. Generally, the performance of GWHP system significantly depends on the geological and hydraulic properties such as hydraulic conductivity, thermal conductivity, soil condition so on. Therefore, in order to use GWHP systems efficiently, it is necessary to analyze the effect of design factors on the system performance. However, there are a few researches on the optimum design method for the open-loop geothermal system. In this research, the design factor in the open-loop geothermal system was analyzed quantitatively for the optimal design method by using numerical simulation. As a result, it was found that the temperature change of heat source depends on the design factor.

• Journal of the Korean Institute of Gas
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• v.4 no.2 s.10
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• pp.7-14
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• 2000

LNG (Liquified Natural Gas) carried by LNG ship is unloaded into the LNG storage tanks at the very low temperature (a little lower than the boiling point of LNG). Because LNG is unloaded through the pipeline, two phase flow appears in the pipeline. In this study, we have studied the pressure-drop mechanisms of the two-phase flow in the pipeline, and the calculation method of BOG (Boil-off Gas) amount based on the heat transfer mechanism through the insulation and the surface of the pipeline. We have developed a computer program for thermal-hydraulic analysis on the LNG pipeline system. We have also developed the optimal design program to find the optimal thickness of insulation and the pipeline size. The program searches the optimal design with the minimum capital cost of pipelines and insulation on the operating conditions of maximum allowance pressure-drop and BOG amount, etc.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.734-743
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• 2017

In order to investigate the swelling behavior and fuel-cladding interaction mechanism of U-10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel-cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal-hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.1100-1108
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• 2017

A thermal-hydraulic system code is an essential tool for the design and safety analysis of a nuclear power plant, and its accuracy quantification is very important for the code assessment and applications. The fast Fourier transform-based method (FFTBM) by signal mirroring (FFTBM-SM) has been used to quantify the accuracy of a system code by using a comparison of the experimental data and the calculated results. The method is an improved version of the FFTBM, and it is known that the FFTBM-SM judges the code accuracy in a more consistent and unbiased way. However, in some applications, unrealistic results have been obtained. In this study, it was found that accuracy quantification by FFTBM-SM is dependent on the frequency spectrum of the fast Fourier transform of experimental and error signals. The primary objective of this study is to reduce the frequency dependency of FFTBM-SM evaluation. For this, it was proposed to reduce the cut off frequency, which was introduced to cut off spurious contributions, in FFTBM-SM. A method to determine an appropriate cut off frequency was also proposed. The FFTBM-SM with the modified cut off frequency showed a significant improvement of the accuracy quantification.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.67-77
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• 2020

Traditionally nuclear thermal hydraulic and nuclear safety has relied on numerical simulations to predict the system response of a nuclear power plant either under normal operation or accident condition. However, this approach may sometimes be rather time consuming particularly for design and optimization problems. To expedite the decision-making process data-driven models can be used to deduce the statistical relationships between inputs and outputs rather than solving physics-based models. Compared to the traditional approach, data driven models can provide a fast and cost-effective framework to predict the behavior of highly complex and non-linear systems where otherwise great computational efforts would be required. The objective of this work is to develop an AI algorithm to predict the peak fuel cladding temperature as a metric for the successful implementation of FLEX strategies under extended station black out. To achieve this, the model requires to be conditioned using pre-existing database created using the thermal-hydraulic analysis code, MARS-KS. In the development stage, the model hyper-parameters are tuned and optimized using the talos tool.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2915-2923
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• 2022

There is an increasing interest in passive safety systems to minimize the need for operator intervention or external power sources in nuclear power plants. Because a passive system has a weak driving force, there is greater uncertainty in the performance compared with an active system. In previous studies, several methods have been suggested to evaluate passive system reliability, and many of them estimated the failure probability using thermal-hydraulic analyses and the Monte Carlo method. However, if the functional failure of a passive system is rare, it is difficult to estimate the failure probability using conventional methods owing to their high computational time. In this paper, a procedure for the application of the Chernoff bound to the evaluation of passive system reliability is proposed. A feasibility study of the procedure was conducted on a passive decay heat removal system of a micro modular reactor in its conceptual design phase, and it was demonstrated that the passive system reliability can be evaluated without performing a large number of thermal-hydraulic analyses or Monte Carlo simulations when the system has a small failure probability. Accordingly, the advantages and constraints of applying the Chernoff bound for passive system reliability evaluation are discussed in this paper.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.9
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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.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1396-1408
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• 2020

Filtered containment venting system (FCVS) is one of the severe accident mitigation systems designed to release containment pressurization to maintain its integrity. The thermal-hydraulic behaviors in FCVSs are important since they affect the operation characteristics of the FCVS. In this study, a representative FCVS was modeled by RELAP5/Mod3.3 code, and the Station BlackOut (SBO) was chosen as an accident scenario. The thermal-hydraulic behaviors of an FCVS during long-term operation with two venting strategies (open-and-close strategy, open-and-non-close strategy) and the sensitivity analysis of important parameters were investigated. The results show that the FCVS can operate up to 250 h with a periodic open-and-close strategy during an SBO. Under the combined effects of steam condensation and water evaporation, the solution inventory in the FCVS increases during the venting phase and decreases during the intermission phase, showing a periodic pattern. Under this condition, the appropriate initial water level is 3-4 m; however, it should be adjusted according to the environment temperature. The FCVS can accommodate a decay heat power of 150-260 kW and may need to feed water for a higher decay heat power or drain water for a lower decay heat power during the late phase. The FCVS can function within an opening pressure range from 450 kPa to 500 kPa and a closing pressure range between 250 kPa and 350 kPa. When the open-and-non-close strategy is adopted, the solution inventory increases quickly in the early venting phase due to steam condensation and then decreases gradually due to the evaporation of water; drying-up may occur in the late venting phase. Decreasing the venting pipe diameter and increasing the initial water level can mitigate the evaporation of the scrubbing solution. These results are expected to provide useful references for the design and engineering application of FCVSs.