• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.996-1007
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• 2019

We investigated the influence of the aspect ratio (H/R) of the oxide layer on the reactor vessel heating in three-layer configuration. Based on the analogy between heat and mass transfers, we performed mass transfer experiments to achieve high Rayleigh numbers ranging from $6.70{\times}10^{10}$ to $7.84{\times}10^{12}$. Two-dimensional (2-D) semi-circular apparatuses having the internal heat source were used whose surfaces of top, bottom and side simulate the interfaces of the oxide layer with the light metal layer, the heavy metal layer, and the reactor vessel, respectively. Multi-cell flow pattern was identified when the H/R was reduced to 0.47 or less, which promoted the downward heat transfer from the oxide layer and possibly mitigated the focusing effect at the upper metallic layer. The top boundary condition greatly affected the natural convection of the oxide layer due to the presence of secondary flows underneath the cold light metal layer.

• KSBB Journal
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• v.16 no.4
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• pp.398-402
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• 2001

Enzymatic hydrolysis of agar was carried out continuously to produce agarooligosaccharides by immobilized agarase in Packed-Bed Reactor. The reactor was constructed using a acryl tube with an internal diameter of 10 mm and a useful height of 140 mm. The Packed-Bed Reactor was 11 mL reactor volume as its length : diameter ratio was 14 : 1. The operation condition of reaction was performed with an 1 g/L agar concentration at 40$^{\circ}C$, 10 mM MOPS buffer(pH 7.0) and with the flow rate 3 mL∼48 mL/h at a dilution rate of 1.09∼5.45 h$\^$-1/. The hydrolysis products was identified DP6, DP4 and DP2 by HPLC. The conversion rate of agar was about 80% and amount of total agarooligosaccharide was 0.88 mg/mL at Packed-Bed Reactor.

• Proceedings of the KSME Conference
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• 2001.06b
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• pp.386-391
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• 2001

In this paper, the vibration characteristics of IHTS(Intermediate Heat Transfer System) piping system of LMR(Liquid Metal Reactor) conveying hot liquid sodium are investigated to eliminate the pipe supports for economic reasons. To do this, a 3-dimensional straight pipe element and a curved pipe element conveying fluid are formulated using the dynamic stiffness method of the wave approach and coded to be applied to any complex piping system. Using this method, the dynamic characteristics including the natural frequency, the frequency response functions, and the dynamic instability due to the pipe internal flow velocity are analyzed. As one of the design parameters, the vibration energy flow is also analyzed to investigate the disturbance transmission paths for the resonant excitation and the non-resonant excitations.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2720-2727
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• 2022

The Generation IV Lead-cooled fast reactor (LFR) take the liquid lead or lead-bismuth eutectic alloy (LBE) as the coolant of the primary cooling circuit. Combined with the natural characteristics of lead alloy and the design features of LFR, the system is the simplest and the number of equipment is the least, which reflects the inherent safety characteristics of LFR. The nuclear main coolant pump (MCP) is the only power component and the only rotating component in the primary circuit of the reactor, so the various operating characteristics of the MCP are directly related to the safety of the nuclear reactor. In this paper, various working conditions that may occur in the normal rotation (positive rotating) of the MCP and the corresponding internal flow characteristics are analyzed and studied, including the normal pump condition, the positive-flow braking condition and the negative-flow braking condition. Since the corrosiveness of LBE is proportional to the fluid velocity, the distribution of flow velocity in the pump channel will be the focus of this study. It is found that under the normal pump condition and positive-flow braking conditions, the high velocity region of the impeller domain appears at the inlet and outlet of the blade. At the same radius, the pressure surface is lower than the back surface, and with the increase of flow rate, the flow separation phenomenon is obvious, and the turbulent kinetic energy distribution in impeller and diffuser domain shows obvious near-wall property. Under the negative-flow braking condition, there is obvious flow separation in the impeller channel.

• International Journal of Advanced Culture Technology
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• v.10 no.3
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• pp.295-306
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• 2022

The NOx removal performance of the SCR process depends on various factors such as catalytic factors (catalyst composition, shape, space velocity, etc.), temperature and flow rate distribution of the exhaust gas. Among them, the uniformity of the flow flowing into the catalyst bed plays the most important role. In this study, the flow characteristics in the SCR reactor in the design stage were simulated using a three-dimensional numerical analysis technique to confirm the uniformity of the airflow. Due to the limitation of the installation space, the shape of the inlet duct was compared with the two types of inlet duct shape because there were many curved sections of the inlet duct and the duct size margin was not large. The effect of inlet duct shape, guide vane or mixer installation, and venturi shape change on SCR reactor internal flow, airflow uniformity, and space utilization rate of ammonia concentration were studied. It was found that the uniformity of the airflow reaching the catalyst layer was greatly improved when an inlet duct with a shape that could suppress drift was applied and guide vanes were installed in the curved part of the inlet duct to properly distribute the process gas. In addition, the space utilization rate was greatly improved when the duct at the rear of the nozzle was applied as a venturi type rather than a mixer for uniform distribution of ammonia gas.

• Nuclear Engineering and Technology
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• v.30 no.1
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• pp.8-16
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• 1998

In this work a method to detect the vibrational peak and to decide the vibrational mode of detected peak for core internal vibration monitoring system which is particularly concerned on the core support barrel (CSB) and fuel assemblies is developed. Flow induced vibration and aging process in the reactor internals cause unsoundness of the internal structure. In order to monitor the vibrational status of core internal, signals from the ex-core neutron detectors are transformed into frequency domain. By analyzing transformed frequency domain signal, an analyst can acquire the information on the vibrational characteristics of the structures, i.e., vibration frequencies of each component, vibrational level, modes of vibration, and the causes of the abnormal vibration, if any. This study is focused on the development of the automated monitoring system. Several methods are surveyed to define the peaks in power spectrum and fuzzy theory is used to automatic detection of the vibrational peaks. Fuzzy algorithm is adopted to define the modes of vibration using the peak values from fuzzy peak recognition, phase spectrum, and coherence spectrum.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.39 no.11
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• pp.905-911
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• 2015

In this study, to examine the effect of a hole size change(smaller hole diameter) in the outer region of the lower-support-structure-bottom plate(LSSBP) on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD software, ANSYS CFX R.15. The predicted results were compared with those of the original LSSBP. Through these comparisons, it was concluded that a more uniform distribution of the mass flow rate at the core-inlet plane could be obtained by reducing the hole size in the outer region of the LSSBP. Therefore, from the nuclear regulatory perspective, design change of the hole pattern in the outer region of the LSSBP may be desirable in terms of improving both the mechanical integrity of the fuel assembly and the core thermal margin.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.71-79
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• 2023

Air ingress incidents are major safety accidents in very-high-temperature reactors (VHTRs). Air containing a high volume fraction of oxygen may cause severe oxidation of core components at the VHTR, especially for the significantly thin alloy tube wall in the intermediate heat exchanger (IHE). The research objects of this study are Inconel 617 and Incoloy 800H, two candidate alloys for IHE in VHTR. The air ingress accident scenario is simulated with high-temperature air flow at 950 ℃. A continuous oxide scale was formed on the surfaces of both the alloys after the experiment. Because the oxide scale of Inconel 617 has a loose structure, whereas that of Incoloy 800H is denser, Inconel 617 exhibited significantly more severe internal oxidation than Incoloy 800H. Further, Inconel 617 showed a significant decrease in ultimate tensile strength and plasticity after aging for 200 h, whereas Incoloy 800H maintained its tensile properties satisfactorily. Through control experiment under vacuum, we preliminarily concluded that serious internal oxidation is the primary reason for the decline in the tensile properties of Inconel 617.

• 한국전산유체공학회:학술대회논문집
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• 2011.05a
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• pp.269-274
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• 2011

A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet($UO_2$) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid 려el by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.

• 한국전산유체공학회:학술대회논문집
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• 2004.03a
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• pp.49-55
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• 2004

This paper addresses a numerical simulation of the flow and heat transfer in a simplified model of helically coiled tube steam generator using a general purpose computational fluid dynamic analysis computer code. The steam generator model is comprised of a cylindrical shell and helically coiled tubes. A cold feed water entered the tubes is heated up, evaporates. and finally become a superheated steam with a large amount of heat transferred continuously from the hot compressed water at higher pressure flowing counter-currently through the shell side. For the calculation of tube side two-phase flow field formed by boiling, inhomogeneous two-fluid model is used. Both the internal and external turbulent flows are simulated using the standard k-e model. The conjugate heat transfer analysis method is employed to calculate the conduction in the tube wall with finite thickness and the convections in the internal and external fluids simultaneously so as to match the fluid-wall-fluid interface conditions properly. The numerical calculations are peformed for helically coiled tubes of steam generator at an integral type pressurized water reactor under normal operation. The effects of tube-side inlet flow velocity are discussed in details. The results of present numerical simulation are considered to be physically plausible based on the data and knowledge from previous experimental and numerical studies where available.