• 한국전산유체공학회:학술대회논문집
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• 2008.03a
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• pp.71-78
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• 2008

For analyses of multi-phase flows in a water-cooled nuclear power plant, a three-dimensional SIMPLE-algorithm based hydrodynamic solver CUPID-S has been developed. As governing equations, it adopts a two-fluid three-field model for the two-phase flows. The three fields represent a continuous liquid, a dispersed droplets, and a vapour field. The governing equations are discretized by a finite volume method on an unstructured grid to handle the geometrical complexity of the nuclear reactors. The phasic momentum equations are coupled and solved with a sparse block Gauss-Seidel matrix solver to increase a numerical stability. The pressure correction equation derived by summing the phasic volume fraction equations is applied on the unstructured mesh in the context of a cell-centered co-located scheme. This paper presents the numerical method and the preliminary results of the calculations.

• 한국전산유체공학회:학술대회논문집
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• 2008.10a
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• pp.71-78
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• 2008

For analyses of multi-phase flows in a water-cooled nuclear power plant, a three-dimensional SIMPLE-algorithm based hydrodynamic solver CUPID-S has been developed. As governing equations, it adopts a two-fluid three-field model for the two-phase flows. The three fields represent a continuous liquid, a dispersed droplets, and a vapour field. The governing equations are discretized by a finite volume method on an unstructured grid to handle the geometrical complexity of the nuclear reactors. The phasic momentum equations are coupled and solved with a sparse block Gauss-Seidel matrix solver to increase a numerical stability. The pressure correction equation derived by summing the phasic volume fraction equations is applied on the unstructured mesh in the context of a cell-centered co-located scheme. This paper presents the numerical method and the preliminary results of the calculations.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.73-78
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• 2010

The suppression of molten salt vaporization is one of the key technical issues in the electrolytic reduction process developed for recycling spent nuclear fuel from light-water reactors Since the Hertz-Langmuir relation previously applied to molten salt vaporization is valid only for vaporization into a vacuum, a diffusion model was derived to quantitatively assess the vaporization of LiCl, $Li_2O$ and Li from an electrolytic reducer operating under atmospheric pressure. Vaporization rates as a function of operation variables were calculated and shown to be in reasonable agreement with the experimental data obtained from thermogravimetry.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.286-300
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• 2002

An algorithm for the sodium boiling model has been developed for calculation of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. Modeling of sodium boiling in liquid metal reactors using sodium as a coolant is necessary because of phenomenon difference comparing with that observed generally in light water reactor systems. The applied model to the algorithm is the multiple-bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbies that (ill the whole cross section of the coolant channel except for the liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble The present study is focused on not only demonstration of the vapor bubble behavior predicted by the developed model, but also confirmation of a qualitative acceptance for the model. As a result, the model can represent important phenomena in the sodium boiling, but it is found that further effort is also needed for its completition.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1381-1389
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• 2019

Stress corrosion cracking (SCC) widely found in both primary and secondary sides of steam generator (SG) tubing in pressurized water reactors (PWR) has become an important safety issue. Using eddy-current tests (ECTs), non-destructive evaluations are performed for the integrity management of SG tubes against intergranular SCC. To enhance the reliability of ECT, this study investigates the effects of oxide films on ECT's detection capabilities for SCC in laboratory-degraded SG tubing in high temperature and high pressure aqueous environment.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3320-3325
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• 2023

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re = 2,409 and Re = 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.637-647
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• 2014

In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite-element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

• Journal of the Korean institute of surface engineering
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• v.56 no.6
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• pp.401-411
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• 2023

This study investigates the synthesis, characterization, and application of zinc oxide (ZnO) nanoparticles for corrosion resistance and stress corrosion cracking (SCC) mitigation in high-temperature and high-pressure environments. The ZnO nanoparticles are synthesized using plasma discharge in water, resulting in rod-shaped particles with a hexagonal crystal structure. The ZnO nanoparticles are applied to Alloy 600 tubes in simulated nuclear power plant atmospheres to evaluate their effectiveness. X-ray diffraction and X-ray photoelectron spectroscopy analysis reveals the formation of thermodynamically stable ZnCr₂O₄and ZnFe₂O₄ spinel phases with a depth of approximately 35 nm on the surface after 240 hours of treatment. Stress corrosion cracking (SCC) mitigation experiments reveal that ZnO treatment enhances thermal and mechanical stability. The ZnO-treated specimens exhibit increased maximum temperature tolerance up to 310 ℃ and higher-pressure resistance up to 60 bar compared to non-treated ZnO samples. Measurements of crack length indicate reduced crack propagation in ZnO-treated specimens. The formation of thermodynamically stable Zn spinel structures on the surface of Alloy 600 and the subsequent improvements in surface properties contribute to the enhanced durability and performance of the material in challenging high-temperature and high-pressure environments. These findings have significant implications for the development of corrosion-resistant materials and the mitigation of stress corrosion cracking in various industries.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.41-48
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• 2016

To investigate the effect of long-term thermal aging on the microstructural and mechanical characteristics of weldment made of nickel base alloy and its weld metal, an accelerated heat treatment was applied to simulate the process of long-term thermal aging in the operating condition of nuclear power plant. A representative nickel-based weldment with Alloy 600 and Alloy 182 was fabricated and heat-treated at $400^{\circ}C$ for 1,713 h and 3,427 h to simulate the thermal aging for the period equivalent to 15 and 30 years in operating pressurized water reactors, respectively. The microstructural and mechanical characteristics were analyzed by using optical microscopy, scanning electron microscopy and Vickers microhardness measurement. Changes were observed in precipitation behavior and microhardness of each specimen, and these changes were mainly attributed to the change in precipitated morphology and residual stress across the weld during the thermal aging process.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.2
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.