• Nuclear Engineering and Technology
• /
• v.55 no.6
• /
• pp.2125-2138
• /
• 2023

The printed circuit heat exchanger (PCHE) with airfoil fins has the benefits of high compactness, high efficiency and superior heat transfer performance. A novel multi-objective optimization approach is presented to design the airfoil fin PCHE in this paper. Three optimization design variables (the vertical number, the horizontal number and the staggered number) are obtained by means of dimensionless airfoil fin arrangement parameters. And the optimization objective is to maximize the Nusselt number (Nu) and minimize the Fanning friction factor (f). Firstly, in order to investigate the impact of design variables on the thermal-hydraulic performance, a parametric study via the design of experiments is proposed. Subsequently, the relationships between three optimization design variables and two objective functions (Nu and f) are characterized by an improved particle swarm optimization-backpropagation artificial neural network. Finally, a multi-objective optimization is used to construct the Pareto optimal front, in which the non-dominated sorting genetic algorithm II is used. The comprehensive performance is found to be the best when the airfoil fins are completely staggered arrangement. And the best compromise solution based on the TOPSIS method is identified as the optimal solution, which can achieve the requirement of high heat transfer performance and low flow resistance.

• Journal of the Korean Geotechnical Society
• /
• v.39 no.11
• /
• pp.53-62
• /
• 2023

The characteristics of unfrozen water content in frozen soils significantly impact the thermal, hydraulic, and mechanical behavior of the ground. A thorough analysis of the unfrozen water content characteristics of the target subsoil material is crucial for evaluating the stability of frozen ground. This study conducted indoor experiments to measure the freezing point and unfrozen water content of sandy soil while considering pore water salinity. Utilizing the experimental data, we introduced a novel empirical model to conveniently estimate the unfrozen water retention curve. Furthermore, the validity of the unfrozen water retention curve was assessed by comparing the experimental data with the results of a simulation model that utilized the proposed empirical model as input data.

• Geomechanics and Engineering
• /
• v.16 no.4
• /
• pp.363-373
• /
• 2018

The evaluation of Thermo-Hydro-Mechanical (THM) coupling behavior is important for the development of underground space for various purposes. For a high-level radioactive waste repository excavated in a deep underground rock mass, the accurate prediction of the complex THM behavior is essential for the long-term safety and stability assessment. In order to develop reliable THM analysis techniques effectively, an international cooperation project, Development of Coupled models and their Validation against Experiments (DECOVALEX), was carried out. In DECOVALEX-2015 Task B2, the in situ THM experiment that was conducted at Horonobe Underground Research Laboratory(URL) by Japan Atomic Energy Agency (JAEA), was modeled by the research teams from the participating countries. In this study, a THM coupling technique that combined TOUGH2 and FLAC3D was developed and applied to the THM analysis for the in situ experiment, in which rock, buffer, backfill, sand, and heater were installed. With the assistance of an artificial neural network, the boundary conditions for the experiment could be adequately implemented in the modeling. The thermal, hydraulic, and mechanical results from the modeling were compared with the measurements from the in situ THM experiment. The predicted buffer temperature from the THM modelling was about $10^{\circ}C$ higher than measurement near by the overpack. At the other locations far from the overpack, modelling predicted slightly lower temperature than measurement. Even though the magnitude of pressure from the modeling was different from the measurements, the general trends of the variation with time were found to be similar.

• Nuclear Engineering and Technology
• /
• v.26 no.1
• /
• pp.1-8
• /
• 1994

The flame speed correlation considering thermal-hydraulic phenomena under severe accidents is proposed and correction coefficients are defined. This correlation modifies the pressure dependency in Iijima-Takeno correlation and adds the steam suppression effects to it in the anticipated hydrogen and steam concentration ranges under severe accidents. The existing models of flame speed due to hydrogen combustion under severe accidents are based on the experiments which were performed merely at room temperature and atmospheric pressure. They have difficulty in predicting a accurate flame speed in a case of high temperature and pressure during severe accidents. Thus the flame structure is assumed as a prerequisite to the reliable determination of flame speed and theoretical model is developed. To examine the validity, flame speeds in various conditions calculated by this model are compared with those obtained by the calculation of the existing correlations of the codes such as improved HECTR and MAAP. Also the steam suppression ratio is quantified and the steam suppression coefficient is defined as a composition of mixture. Initial temperature and pressure dependencies are investigated and correction coefficents are determined. More experimental studies can be recommended to improve this correlation to its further works.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.24 no.5
• /
• pp.744-752
• /
• 2000

The influence of arrangement and length of discrete ribs on heat/mass transfer and friction loss is investigated. Mass transfer experiments are conducted to obtain the detailed local heat/mass transfer information on the ribbed wall. The aspect ratio (width/height) of the duct is 2.04 and the rib height is one tenth of the duct height, such that the ratio of the rib height to hydraulic diameter is 0.0743. The ratio of rib-to-rib distance to rib height is 10. The discrete ribs were made by dividing each continuous rib into 2, 3 or 5 pieces and attached periodically to the top and the bottom walls of the duct with a parallel orientation The combined effects of rib angle and length of the discrete ribs on heat/mass transfer ae considered for the rib angles $({\alpha})\;of\;90^{\circ}\;and\;45^{\circ}$. As the number of the discrete ribs increases, the uniformity of the heat/mass transfer distributions increases. For $(\alpha})=90^{\circ}$, the heat/mass transfer enhancement with the discrete ribs is remarkable, while the heat/mass transfer performances are slightly higher than that of the transverse continuous ribs due to the accompanied high friction loss penalty. For $(\alpha})=90^{\circ}$, the average heat/mass transfer coefficients and the heat/mass transfer performances decrease slightly with the discrete ribs compared to the case of the angled continuous ribs.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.40 no.7
• /
• pp.457-470
• /
• 2016

Most CFD codes, that mainly adopt the heat partitioning model as the wall boiling model, have shown low accuracies in predicting the two-phase flow parameters of subcooled boiling phenomena under low pressure conditions. In this study, a number of subcooled boiling experiments in vertical channels were analyzed using a thermal-hydraulic component code, CUPID. The prediction of the void fraction distribution using the CUPID code agreed well with experimental data at high-pressure conditions; whereas at low-pressure conditions, the predicted void fraction deviated considerably from measured ones. Sensitivity tests were performed on the submodels for major parameters in the heat partitioning model to find the optimized sets of empirical correlations suitable for low-pressure subcooled flow boiling. The effect of the K-factor on the void fraction distribution was also evaluated.

• Journal of Energy Engineering
• /
• v.21 no.4
• /
• pp.419-426
• /
• 2012

A transient, three-dimensional, two-phase flow analysis code, CUPID, has been developed in KAERI. In this work, we performed a preliminary analysis using the CUPID code to investigate the thermal-hydraulic behavior of the moderator in the Calandria vessel of a CANDU reactor. At first, we validated the CUPID code using the three experiments that were performed at Stern Laboratories Inc. To avoid the complexity to generate computational mesh around the Calandria tube bundles, a porous media approach was applied for the region. The pressure drop in the porous media zone was modeled by an empirical correlation. The results of the calculations showed that the CUPID code can predict the mixed flow pattern of forced and natural convection inside the Calandria vessel very well. Thereafter, the analysis was extended to a two-phase flow condition. Also, the local maximum temperature in the Calandria vessel was plotted as a function of the injection flow rate, which may be utilized to predict the local subcooling margin.

• Journal of the Korean Society of Radiology
• /
• v.11 no.7
• /
• pp.571-578
• /
• 2017

Bubbles are generated by the boiling of the cooling water when an accident occurs in the reactor and then in order to measure the void fraction, the Optical Fiber Probe(OFP) and optical camera are used in thermal hydraulic safety research. However, such an optical method is not suitable for measuring the void fraction in a $17{\times}17$ array of fuel rods due to the geometrical limitations. This study was conducted as a preliminary study using x-ray system and various phantoms before applying to rod bundles. Through radiographic and tomographic experiments, the tube voltage of the x-ray generator was 130 kVp and the tube current was 1 mA. In addition, it is possible to measure the hole of 1mm in size visually through the bubble resolution phantom, and it is confirmed that the contrast is relatively decreased in the inside of the freon in the case of the contrast evaluation using the road phantom. However, we could obtain good image without distortion when reconstructing the image. Bubble generation phantom experiments were used to confirm the flow direction of the bubbles and to acquire tomography images. The image J tool was used to measure the void fraction of 18 % for a single tomography image. This study has carried out previous researches for the measurement of the bubble rate around the nuclear fuel and could be used as a basic research for continuous research.

• Nuclear Engineering and Technology
• /
• v.49 no.1
• /
• pp.54-70
• /
• 2017

The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility-the advanced core-cooling mechanism experiment (ACME)-was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups-a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break-were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative effects on the passive core cooling performance caused by nitrogen injection during the SBLOCA transient.

• Journal of Energy Engineering
• /
• v.25 no.2
• /
• pp.1-20
• /
• 2016

A condensation heat transfer model is very important for the safety analysis of nuclear power plants. Especially, condensation under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NCGs in the vapor largely reduces the condensation rate. In this study, the condensation heat transfer model of the severe accident analysis code MELCOR 1.8.6 has been assessed using a set of condensation experiments performed under the thermal-hydraulic conditions similar to those inside a containment during design-basis accidents or severe accidents. Experiment conditions are categorized into 4 types according to the shape of the condensation surface: vertical flat plates, outer surface of vertical pipes, inner surface of vertical pipes, the inner surface of horizontal pipes. The results of the calculations show that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on inner surface of vertical pipes.