• Proceedings of the KSME Conference
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• 2003.04a
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• pp.585-590
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• 2003

The distributions of mass flow rate and pressure are major factors to deside the performance of a proton exchange membrane fuel cell (PEMFC). These factors are affected by channel configuration of air plate. In this paper. structural analysis is performed to investigate deformation of porous media. Two kind of models are suggest for flow analyses. Deformed porous media and undeformed porous media are considered for air plate model. The Numerical flow analysis results with deformed porous media and undeformed porous media had some discrepancy in pressure distribution. The pressure and velocity in a working condition are numerically calculated to predict the performance of the air plates. Distributions of the parameters in the PEMFC are analyzed numerically under steady-state conditions.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.57-61
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• 2008

During the period of retubing work for the licensing renewal, the fuel channels, calandria tubes and feeders of CANDU Reactors will be replaced, and calandria visual examination will be performed. This period is a unique opportunity to inspect the inside of the calandria. The visual inspection for the calandria vessel and its internals of Wolsong NPP is scheduled for confirming the calandria integrity. The first visual inspection for the calandria is planned in Pt. Lepreau led by AECL. The visual inspection for Wolsong NPP, led by NETEC(Nuclear Engineering & Technology Institute) of KHNP, will employ 3D laser scanner and 3D CAD Mock-up for the first time in the world, in addition to a conventional video camera. The inspection system is composed of a robot with the 3D laser scanner, a video camera and a hardness meter.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.915-921
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• 2019

Radiation damage due to neutrons has been calculated in Ni-based alloys in Wolsong CANDU reactor environments. Two damage parameters are considered: displacement damage, and transmutation gas production. We used the SPECTER and SRIM computer codes in quantifying radiation damage. In addition, damage caused by Ni two-step reactions was considered. Estimations were made for the annulus spacers in a CANDU reactor that are located axially along a fuel channel and made of Inconel X-750. The calculation results indicate that the transmutation gas production from the Ni two-step reactions is predominant as the effective full power year increases. The displacement damage due to recoil atoms produced from Ni two-step reactions accounts for over 30% out of the total displacement damage.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2222-2227
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• 2007

Newly structured metal-supported solid oxide fuel cell was fabricated and characterized by impedance analysis and galvanodynamic experiment. Using a cermet adhesive, thin ceramic layer composed of anode(Ni/YSZ) and electrolyte(YSZ) was joined with STS430 metal support of which flow channel was fabricated. $La_{0.8}Sr_{0.2}Co_{0.4}Mn_{0.6}O_3$ perovskite oxide was used as cathode material. Single cell performance was increased and saturated at operating time to 300hours at 800$^{\circ}C$ because of cathode sintering effect. The sintering effect was reinvestigated by half cell test and exchange current density was measured as 0.005A/$cm^2$. Maximum power density of the cell was 0.09W/$cm^2$ at 800$^{\circ}C$. Numerical analysis was carried out to classify main factors influencing the single cell performances. Compared to experimental IV curve, simulated curve based on experimental parameters such as exchange current density was in good agreement.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2648-2653
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• 2007

Recently, increasing demand on not only lighter but also extremely mobile battery make micro fuel cell device very attractive alternative. By reducing the size of fuel cell, surface tension becomes dominant factor with minor gravitational effect. Therefore, it is very difficult to detach the $CO_2$ bubble generating on a cathode side in ${\mu}DMFC$ (micro direct methanol fuel cell). The degassing of a $CO_2$ bubble has drawn quite attention especially for ${\mu}DMFC$ due to its considerable effect on overall machine performance. Our attention has been paid to the dynamic behavior of immiscible bubble attached to the one side of the wall on 2D rectangular channel subject to external shear flow. We use Level Contour Reconstruction Method (LCRM) which is simplified version of front tracking method to track the bubble interface motion. Effects of Reynolds number, Weber number, advancing/receding contact angle and property ratio on bubble detachment characteristic has been numerically identified.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.292-295
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• 2008

In a SCWR (SuperCritical pressure Water cooled Reactor), the coolant temperature initially at below the pseudo-critical temperature at the bottom of a reactor core increases as the coolant flows upward through the sub-channels of the fuel assemblies, and it finally becomes higher than the pseudo-critical temperature when it leaves the reactor core. At certain conditions, heat transfer deterioration occurs near the pseudo-critical temperature and it may cause a drastic rise of the fuel surface temperature resulting a fuel failure. Therefore, an accurate estimation of the heat transfer coefficient is very important for the thermal-hydraulic design of a reactor core. An experiment on heat transfer to the vertically upward flowing $CO_2$ at a supercritical pressure in a circular tube were performed at KAERI. The internal diameter of the test section is 6.32 mm, which corresponds to the hydraulic diameter of a sub-channel in the conceptional design proposed by KAERI. The test range of the mass flux is 285 to 1200 kg/m$^2$s and the maximum heat flux is 170 kW/m$^2$. The inlet pressure is maintained at 8.12 MPa, which is 1.1 times the critical pressure. A new correlation, which covers both the normal and deterioration heat transfer regimes was proposed and compared with the estimations by exiting correlations.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2007.04a
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• pp.148-151
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• 2007

We made two injectors that were equal to all design except for existence or nonexistence of swirl chamber of fuel part, because we want to find cooling characteristics at the injector face according to existence or non existence of swirl chamber of fuel part. And we set regenerative cooling channel in injector face for protecting injector face for prolonged combustion time. Two injectors were performed hot firing test, and then we compared cooling characteristics of two injectors. Also we compared O/F ratio effects on cooling characteristics and combustion characteristics.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.687-700
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• 1995

The thermal-hydraulic transient analysis of the proposed CANDU-9 plant was peformed. Several major transients ore analyzed if they meet the heat transport system design requirements. The proposed heat transport system configuration and the preliminary sizes of system equipment are justified by analysis in terms of the fuel integrity and the high system pressure limit during transients. The compliance with AECB R-77 requirements for CANDU-9 reactor was estimated. The analysis results showed that for each postulated accident the peak pressure values in the reactor headers are within the acceptance criteria given in ASME code requirements and the fuel overheating is prevented. One pump start-up during the reactor start-up operation was analyzed to investigate the How reversal through the fuel channel, which is specific in the proposed CANDU-9 plant.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.581-588
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• 2013

This paper deals with the Safety Analysis for $CANDU^{(R)}$ 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.