• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3754-3767
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• 2023

Turbulent jet discharges in subcooled water pools are essential for safety systems in nuclear power plants, specifically in the pressure suppression pool of boiling water reactors and In-containment Refueling Water Storage Tank of advanced pressurized water reactors. The gas and liquid flow in these systems is investigated using multiphase flow analysis. This field has been extensively examined using a combination of experiments, theoretical models, and Computational Fluid Dynamics (CFD) simulations. ANSYS CFX offers two approaches to model multiphase flow behavior. The non-homogeneous Eulerian-Eulerian Model has been used in this work; it computes global information and is more convenient to study interpenetrated fluids. This study utilized the Large Eddy Simulation Model as the turbulence model, as it is better suited for non-stationary and buoyant flows. The CFD results of this study were validated with experimental data and theoretical results previously obtained. The figures of merit dimensionless penetration length and the dimensionless buoyancy length show good agreement with the experimental measurements. Correlations for these variables were obtained as a function of dimensionless numbers to give generality using only initial boundary conditions. CFD numerical model developed in this research has the capability to simulate the behavior of non-condensable gases discharged in water.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2276-2296
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• 2022

This study validates the applicability of the CUPID code for simulating subcooled wall boiling under high-pressure conditions against number of DEBORA tests. In addition, a new numerical technique in which the interfacial momentum non-drag forces are calculated at the cell faces rather than the center is presented. This method reduced the numerical instability often triggered by calculating these terms at the cell center. Simulation results showed good agreement against the experimental data except for the bubble sizes in the bulk. Thus, a new model to calculate the Sauter mean diameter is proposed. Next, the effect of the relationship between the bubble departure diameter (D_dep) and the nucleation site density (N) on the performance of the Wall Heat Flux Partitioning (WHFP) model is investigated. Three correlations for D_dep and two for N are grouped into six combinations. Results by the different combinations show that despite the significant difference in the calculated D_dep, most combinations reasonably predict vapor distribution and liquid temperature. Analysis of the axial propagations of wall boiling parameters shows that the N term stabilizes the inconsistences in D_dep values by following a behavior reflective of D_dep to keep the total energy balance. Moreover, ratio of the heat flux components vary widely along the flow depending on the combinations. These results suggest that separate validation of D_dep correlations may be insufficient since its performance relies on the accompanying N correlations.

• Nuclear Engineering and Technology
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• v.51 no.6
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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.

• Journal of the korean Society of Automotive Engineers
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• v.10 no.5
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• pp.69-82
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• 1988

Heat transfer experiment is carried out during the performance test of the 4-cylinder 4-stroke cycle turbo-charged gasoline engine. Cycle simulation employing the measured pressure in cylinder, the cooling water temperature and flow rate and others is carried out in order to calculate the gas temperature in cylinder. In this simulation combustion process was simulated by Annand's two zone model and suction, compression, and other processes are calculated completely. From this simulation, we can obtain not only the heat transfer coefficient but also the flame speed, turbulent burning velocity, flame factor and the boiling condition of cooling passage. The results are investigated with engine speed, equivalence ratio and spark advance.

• Aerospace Engineering and Technology
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• v.9 no.2
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• pp.193-197
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• 2010

Water cooling ducts are installed in the exhaust diffuser for high altitude tests of rocket engine to protect diffuser from high-temperature combustion gas. The mass flow rate and pressure of cooling water is designed to prevent boiling of cooling water in the ducts. Therefore, the estimation of maximum temperature of duct wall is important parameter in design of cooling system, especially pressure of cooling water. The method for predicting maximum temperatures of duct walls with variation of coolant flow rates was derived theoretically.

• Journal of Energy Engineering
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• v.25 no.1
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• pp.69-75
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• 2016

Bubble size is a key parameter for an accurate prediction of bubble behaviours in the multi-dimensional two-phase flow. In the current STAR CCM+ CFD code, a mechanistic bubble size model $S{\gamma}$ is available for the prediction of bubble size in the flow channel. As another model, Yun model is developed based on DEBORA that is subcooled boiling data in high pressure. In this study, numerical simulation for the gas-liquid two-phase flow was conducted to validate and confirm the performance of $S{\gamma}$ model and Yun model, using the commercial CFD code STAR CCM+ ver. 10.02. For this, local bubble models was evaluated against the air-water data from DEDALE experiments (1995) and Hibiki et al. (2001) in the vertical pipe. All numerical results of $S{\gamma}$ model predicted reasonably the two-phase flow parameters and Yun model is needed to be improved for the prediction of air-water flow under low pressure condition.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 2003.10a
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• pp.281-283
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• 2003

High temperature superconducting power cable requires forced flow cooling. Liquid nitrogen is circulated by a pump and cooled back by cooling system. Typical operating temperature range is expected to be between 65K and 80K. Subcooler heat exchanger uses saturated liquid nitrogen boiling on the shell side to subcool the circulating liquid nitrogen stream that cools the HTS cable. The paper describes performance tests of the cooling system. The test items are heat exchanging performance of subcooler. pressure drop between supply and return lines, heat transfer coefficient inside former, cable cryostat heat leak and simulation of electrical load of HTS cable.

• The KSFM Journal of Fluid Machinery
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• v.14 no.3
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• pp.10-17
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• 2011

Free energy based lattice Boltzmann method (LBM) has been used to simulate the contact angle and the bubble necking with large density ratio. LBM with the proper contact angle model is able to reduce the spurious currents and eliminate the singularity in the contact lines. The numerical results of the contact angles are satisfied with the Youngs law. For bubble necking flows, simulations are executed for various viscosities and contact angles. The phenomena of the bubble necking are simulated successfully and the subsequent results are presented. The present method is also applicable to the nucleate boiling flows.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.111-120
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• 2021

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great influence on the cooling ability of fuel debris in low plenum, it's very necessary to evaluate the transient changes of the shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a large number of debris bed relocation experiments were carried out by improved bottom gas-injection experimental method in this paper. The effects of different experimental factors on the relocation process were studied from the experiments. The experimental data were also used to further evaluate a semi-empirical onset model for predicting relocation.