• 설비공학논문집
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• 제20권8호
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• pp.541-546
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• 2008

A coolant operating condition in al fuel cell stack was an important factor to determine the temperature distribution which affected the fuel cell performance and relative humidity. In this study, the fuel cell performance was evaluated as a function of the coolant flow rate with the range of $0.1{\sim}0.8$ liter/min cell and the coolant inlet temperature of $20{\sim}82^{\circ}C$. The cell temperature increased with increasing the coolant inlet temperature and with decreasing the coolant flow rate. The coolant inlet temperature and flow rate to maintain the better performance of the fuel cell were in the range of $45{\sim}60^{\circ}C$ and $0.2{\sim}0.4$ liter/min cell, respectively. The experimental results showed that the optimal heat removal rate from the stack by coolant was $0.4{\sim}0.6W/cm^2$ cell.

• 설비공학논문집
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• 제7권2호
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• pp.319-328
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• 1995

In this study, the experiment with 2rows-2columns fin-tube heat exchanger under forced convection and frosting condition is performed. The influence of each operating condition(the temperature of air, the humidity of air, the velocity of air, the temperature of coolant) on the growth of frost layer, air-side pressure drop, and characteristics of heat transfer is investigated. The experimental results show that the frost thickness increases rapidly in the early stage of frost formation and increases linearly after sometime. The frost thickness increases with the increase of the inlet air humidity and velocity and the decrease of inlet air temperature and coolant temperature. It is also found that the total energy transfer rate increases with the increase of inlet air temperature and velocity and with the decrease of inlet air humidity and coolant temperature.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.71-81
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• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

• 반도체디스플레이기술학회지
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• 제20권4호
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• pp.134-140
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• 2021

This study investigates the flow efficiency and temperature based on flow path shape. Five models are designed to the no flow path, one flow path, two flow path, three flow path, add inlet flow path and add interior space gradient. Results show that two flow model(add inlet flow path and add interior space gradient), It was confirmed that model(add inlet flow path) is the optimal shape for coolant heat transfer, and model(add interior space gradient) is the optimal shape for coolant flow, demonstrates optimal design among the five models. The results of this study can be utilized to efficiently control the coolant flow through various types of flow paths.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1532-1539
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• 2019

This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2048-2054
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• 2022

The coolant temperature feedback coefficient is an important parameter of reactor core power control system. To study the coolant temperature feedback coefficient influence on the core power control system of small PWR, the core power control system is built with the nonlinear model and fuzzy control theory. Then, the uncertainty quantification method of reactor core parameters is established based on the Latin hypercube sampling method and the Bootstrap method. Finally, under the conditions of reactivity step perturbation and coolant inlet temperature step perturbation, uncertainty analysis for two cases is carried out. The result shows that with fuzzy controller and fuzzy PID controller, the uncertainty of the coolant temperature feedback coefficient affects the core power control system, and the maximum uncertainties of core relative power, coolant temperature deviation, fuel temperature deviation and total reactivity are acceptable.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.941-951
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• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• Nuclear Engineering and Technology
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• 제29권3호
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• pp.186-195
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• 1997

When asymmetric thermal-hydraulic conditions occur between cold legs, the core inlet temperature will be nonuniform if the coolant is not mixed perfectly in the lower plenum. These uneven core inlet conditions may induce the change in core power distribution. Thus realistic prediction of thermal mixing is important in such abnormal conditions. In this study, reactor internals, which are scaled down as to conserve the flow area ratio, are set up in the model of KORI Unit 1 with the scaling factor of 1/710 by volume and coolant temperatures are measured beneath the lower core plate. Based on experimental results, the ability of COMMIX-1B code to simulate the coolant mixing phenomena in the lower plenum is estimated. The results show that complete mixing never occurs in any conditions and the mixing pattern is characterized according to the plant type.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1747-1755
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• 2021

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

• 동력기계공학회지
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• 제15권4호
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• pp.42-47
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• 2011

The purpose of this study is to investigate the best melting condition with various winding number of a heating pipe, supplying quantity of engine coolant and coolant temperature at the inlet of the heating pipe. Also, it is to suggest getting method of an appropriate quantity of the agent for the urea-SCR system within 10 minutes. For this matter, this study identifies the temperature distribution of inside of urea-tank while it is frozen at the low temperature condition, and suggests the best melting condition of the frozen urea within 10 minutes. From the results, it was found that 2L of melted urea was obtained by the coolant flow rate of 200L/hr at $70^{\circ}C$ for 10 minutes from the start of engine operating.