• Solar Energy
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• v.4 no.2
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• pp.43-49
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• 1984

Experiments were performed to study freezing on a finned vertical tube when either conduction in the solid or natural convection in a liquid controls the heat transfer. Conduction is the controlling mode when the liquid is at its fusion temperature, whereas natural convection controls when the liquid temperature is above the fusion value. The liquid was housed in a cylinderical containment vessel whose surface was maintained at a uniform, time-invariment temperature during a data run, and the freezing occurred on a finned vertical tube positioned along the axis of the vessel. The phase change medium was n-octacosan, a paraffin which freezes at about $61^{\circ}C$. For conduction-controlled freezing, the enhancement of the frozen mass due to finning is greatest when the frozen layer is thin and decrease as the layer grows thicker. The degree of enhancement is generally less than the surface area ratio of the finned and unfinned tube.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.14 no.12
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• pp.1063-1070
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• 2002

To resist ice adhesion on cooling wall is concerned to continuous ice formation in thermal storage system. In this study, ice slurries were formed with two ecological aqueous solution, one is ethanol+silanol and the other is propylene glycol+silanol. By freezing under stirring the solution of $300m\ell$ in a stainless steel vessel which was immersed and cooled in a temperature controlled bath, the shape of ice slurry and the strength of ice adhesion on wall was observed with measuring the temperature and stirring load variation. As the concentration is smaller and the supercooling degree is larger, the ice adhesion is easy to occur. When the stirring load is larger than$ 2.1\times10^{-5}W$, the ice adhesion occurred.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.207-216
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• 2014

This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accident management measures at the Fukushima Daiich nuclear power plants seem to be not fully effective, limitations of current severe accident management strategy are discussed to identify the areas for the potential improvements including core cooling strategy, containment venting, hydrogen control, depressurization of primary system, and proper indication of event progression. The gap between the Fukushima accident event progression and current understanding of severe accident phenomenology including the core damage, reactor vessel failure, containment failure, and hydrogen explosion are discussed. Adequacy of current safety goals are also discussed in view of the socio-economic impact of the Fukushima accident. As a conclusion, it is suggested that an investigation on a coherent integrated safety principle for the severe accident and development of innovative mitigation features is necessary for robust and resilient nuclear power system.

• Journal of Ocean Engineering and Technology
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• v.18 no.1
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• pp.69-74
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• 2004

In general, the protection method of Shell and Tube Type heat exchanger for a vessel has been applied as a sacrificial anode, which is attached at the inner side of the shell. However, this is an insufficient protection method for tube. Therefore, a more suitable method, such as the impressed current cathodic protection for tube protection, is required. Al-brass is the raw material of tubes for heat exchanger of a vessel where seawater is used for cooling the water. It has a high level of heat conductivity, excellent mechanical properties, and a high level of corrosion resistance, due to a cuprous oxide (Cu$_2$O) layer against th seawater. However, in actuality, it has been reported that Al-brass tubes for heat exchanger of a vessel can produce local corrosion, such as stress corrosion cracking (SCC). This paper studied the effect of impressed potential on the stress corrosion cracking of Al-brass for impressed current cathodic protection in 3.5% NaCl +0.1% NH$_4$OH solution, under flow by a constant displacement tester. Based on the test results, the latent time of SCC, stress corrosion crack propagation, and the dezincification phase of Al-brass are investigated.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2188-2197
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• 2022

To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it is regulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtained by the internal pressure and thermal load should be obtained through crack analysis of the nozzle corner crack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G, the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling or heating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomial format based on the stress linearization analysis results. Inevitably, the SIF can only be obtained through finite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal load under, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant, three geometric variables were set and 72 cases of RPV models were made, and then the heat transfer analysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions are suggested.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.14 no.12
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• pp.1071-1077
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• 2002

Ice adhesion on cooling wall is very important in continuous ice formation. The purpose of the present study is to investigate the possibilities of a three-component aqueous solution as a thermal storage material for the continuous ice formation. By freezing under stirring the solution of 300 mL in a stainless steel vessel which was immersed and cooled in a temperature controlled bath, an ice slurry was formed experimentally with measuring the temperature and stirring load variation. From the experiment, the ice adhesion was suppressed when the supercooling degree decreased and the concentration of aqueous solution increased.

• Journal of Advanced Marine Engineering and Technology
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• v.32 no.2
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• pp.299-305
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• 2008

This present study has dealt with the natural convection of water in a rectangular vessel which has cooling point at the center of itself with numerically. The finite difference method (FDM) is presented for the two-dimensional computer simulation of water controlled by natural convection and heat conduction. According to this study, It is cleared that the overturn of density is clearly existed at the temperature of $4[^{\circ}C]$ and that was compared with experimental result. Also the change of natural convection is known from the streamlines and isotherms. Most of all. It is cleared that the overturn of natural convection is changed with time caused by the fact that the temperature and density relationship of water.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.3 no.1
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• pp.44-48
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• 2002

It is necessary to keep Kimchi at 0~5$^{\circ}C$ all around the year to maintain its taste, which is very sensitive to temperature changes. In addition to temperature for maintenance of taste, property of container is also important to control the activity of bacteria involved in Kimchi. As a pottery container is the best one for the bacterial activity, it was used to store Kimchi. Pottery is shielded by cooling pipe directly, and temperature of Kimchi is controlled by the independent temperature controller.