• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1603-1610
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• 2020

Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1109-1118
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• 2021

The construction of STELLA-2 facility is on-going to demonstrate the safety system of PGSFR and to provide comprehensive understanding of transient behavior under DBEs. Considering that most events are single-phase natural circulation flow with slow transient, STELLA-2 was designed with reduced-height of 1/5 length scale. The ratio of volume to surface area in the vessel can relatively increase resulting in excessive heat transfer. Therefore, a steady-state thermal-hydraulic analysis was performed and the effect of design change to reduce the heat transfer through redan was investigated. The heat transfer through single wall redan in STELLA-2 was 3% of the core power, comparable to 1% of the core power in PGSFR. By applying the insulated redan, about 70% of decrease effect was observed. The effect on transient behavior was also evaluated. The conclusion of this study was directly applied to the STELLA-2 design and the modified version is under construction.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.353-363
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• 2023

The STELLA-2 is a large-scale sodium thermal-hydraulic integral effect test facility and supports the development of PGSFR. The facility adopted Pump Simulation Loop System (PSLS) concept for the mechanical sodium pump in the reference reactor to control and to measure the primary sodium flow. Since the component (mechanical pump) is replaced by the loop, it is very important to evaluate the similarity between the pump and the loop. In this paper, to simulate the characteristic of the mechanical sodium pump, the pressure loss along the various options of the loop was evaluated and the comprehensive validity of each design options was analyzed. Using the similarity criteria based on the Richardson number and Euler number conservation, the PSLS design was finalized and the result was within the acceptable error range. Finally, the result of this study was used for construction of the overall facility, STELLA-2.

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

• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.340-351
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• 1991

Experimental studies were conducted on a l/5.03 scale reactor flow model of the Yong-gwang Nuclear Units 3 and 4. The purpose of the flow model test was to estimate the hydraulic effect in the reactor vessel due to the relative size difference between the ABB-CE's System 80 and the YGN 3&4 reactors. The flow model was designed according to the principle of similarity. Obtained from the test were the core inlet flow distribution, the core exit pressure deviations, and the segmental and overall pressure losses across the flow path from the reactor vessel inlet to outlet nozzle. These data will be used to provide input data for the core thermal margin analysis and to verify the analytical hydraulic design method.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.599-604
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• 1997

One of important inconsistencies in the six-equation model predictions has been found to be the force experienced by a single bubble placed in a convergent stream of liquid. Various sets of governing equations yield different amount of forces to hold the bubble stationary in a convergent nozzle. By using the first order potential flow theory, it is found that the six-equation model can not be used to estimate the force experienced by a deformed bubble. The theoretical value of the particle stress of a bubble in a convergent nozzle flow has been found to be a function of the Weber number when bubble distortion is allowed. This force has been calculated by using different sets of governing equations and compared with the theoretical value. It is suggested in this study that the bubble size distribution function can be used to remove the presented inconsistency by relating the interfacial variables with different moments of the bubble size distribution function. This study also shows that the inconsistencies in the thermal-hydraulic governing equation can be removed by mechanistic modeling of the phasic interface.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.237-246
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• 1994

A high flow critical heat flux (CHF) correlation, based on the single-pin CHF experimental data for finned and unfinned heated rods, was developed for the thermal-hydraulic design and safety analysis of the Korea Multi-purpose Research Reactor (KMRR) core. The correlation consists of dimensionless parameters such as Reynolds number, thermodynamic equilibrium quality, liquid-to-vapor density ratio, and hydraulic equivalent diameter ratio. The fin effect was taken into account in the correlation by a finned-to-unfinned heated perimeter ratio. The effects of a cold wall and non-uniform axial power distribution ore discussed to verify the applicability of the single-pin based correlation to the KMRR fuel bundle. The correlation limit departure from nucleate boiling ratio (DNBR) was determined as 1.44 from the statistical analysis of the CHF data.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.26 no.8
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• pp.1153-1163
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• 2002

A transient analysis on temperatures of fuel and oil in hydraulic and lubrication systems in an aircraft was studied using the finite difference method. Numerical calculation was performed by an explicit method with modified Dufort-Frankel scheme. Among various missions, air superiority mission was considered as a mission model with 20% hot day ambient condition in subsonic region. The ambience of the aircraft was assumed as turbulent flow. Convective heat transfer coefficient were used in calculating heat transfer between the aircraft surface and the ambience. For an aircraft on the ground, an empirical equation represented as a function of free-stream air velocity was used. And the heat transfer coefficient for flat plate turbulent flow suggested by Eckert was employed for in-flight phases. The governing equations used in this analysis are the mass and energy conservation equations on fuel and oils. Here, analysis of fuel and oil temperature in the engine was not carried out. As a result of this analysis, the ground operation phase has shown the highest temperature and the largest rate of temperature increase among overall mission phases. Also, it is shown that fuel flow rate through fuel/oil heat exchanger plays an important role in temperature change of fuel and oil. This analysis could be an important part of studies to ensure thermal stability of the aircraft and can be applicable to thermal design of the aircraft fuel system.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.8
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• pp.831-836
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• 2014

In this study, high temperature integrity evaluation on a pressure vessel of the expansion tank operating at elevated temperature of $510^{\circ}C$ in the sodium test facility of the SEFLA(Sodium Thermal-hydraulic Experiment Loop for Finned-tube Sodium-to-Air heat exchanger) to be constructed at KAERI has been performed. Evaluations of creep-fatigue damage based on a full 3D finite element analyses were conducted for the expansion tank according to the recent elevated temperature design codes of ASME Section III Subsection NH and French RCC-MRx. It was shown that the expansion tank maintains its integrity under the intended creep-fatigue loads. Quantitative code comparisons were conducted for the pressure vessel of austenitic stainless steel 316L.

• Journal of Energy Engineering
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• v.28 no.4
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• pp.103-110
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• 2019

The domestic innovative power reactor named iPOWER will employ the passive molten corium cooling system(PMCCS) to cool down and stabilize the core melt in the severe accident. The final design concept of the PMCCS is yet to be determined, but the in-vessel retention through external reactor vessel cooling has been also considered as a viable strategy to cope with the severe accident. In this study, the two-phase natural circulation flow established between the reactor vessel and the insulation was simulated using a thermal-hydraulic system code, MARS-KS. The flow path of cooling water was modeled with one-dimensional nodes, and the boundary condition of the heat load from the molten core was defined to estimate the naturally-driven flow rate. The evolution of major thermal-hydraulic parameters were also evaluated, including the temperature and the level of cooling water, the void fraction around the lower head of the reactor vessel, and the heat transfer mode on its external surface.