• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.12
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• pp.1069-1078
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• 2013

Recently, condensation heat exchangers have been studied for applications to the passive cooling systems of nuclear plants. To design vertical-type condensation heat exchangers in secondary passive cooling systems, TSCON (Thermal Sizing of CONdenser), a thermal sizing program for a condensation heat exchanger, was developed at KAERI (Korea Atomic Energy Research Institute). In this study, the existing condensation heat transfer correlation of TSCON was evaluated using 1,157 collected experimental data points from the heat exchanger of a secondary passive cooling system for the case of pure steam condensation. The investigation showed that the Shah correlation, published in 2009, provided the most satisfactory results for the heat transfer coefficient with a mean absolute error of 34.8%. It is suggested that the Shah correlation is appropriate for designing a condensation heat exchanger in TSCON.

• Journal of the Korean Society of Safety
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• v.20 no.1 s.69
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• pp.42-48
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• 2005

The present study investigates the effect of balcony on external smoke movement of high rise building through the fire tests of the 1/10 reduced model scale using Froude scaling. A hexane pool fire is used to examine the smoke movement for various opening sizes of balcony and temperature distributions are measured by T-type thermocouples. Also, hydrogen bubble technique is applied to visualize the smoke movement near the balcony. Measured temperatures of the closed balcony is 2.5 times higher than those of the open balcony because the external smoke in case of the closed balcony rise along the vertical wall. The maximum vertical temperature of partially closed balcony is similar with fully closed balcony and mean temperature inside of balcony increases as opening size of balcony decreases. The experimental results show that the balcony space plays an important roles in preventing fire propagation and cooling of smoke layer. In order to ensure the fire safety in high rise building design, a series of systematic researches are required to examine the various type of balconies.

• Nuclear Engineering and Technology
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• v.17 no.1
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• pp.16-24
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• 1985

A computer model is developed capable of simulating the transient behavior of a pool-type liquid metal-cooled fast breeder reactor (LMFBR). The model, SIMFARP, is a fast running computer code which may be used to simulate the loss of power to any pump(s), a complete loss-of-forced cooling, and the natural circulation behavior. Eight governing equations are derived and a Runge-Kutta algorithm is applied to integrate the eight differential equations. The developed computer program is applied to two cases; loss of electric power to any pump(s), and loss of all external electric supply power without scram in Super-Phenix-I.

• Tunnel and Underground Space
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• v.15 no.1 s.54
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• pp.55-60
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• 2005

In this study, the 1/20 reduced-scale experiments using Froude scaling were conducted to investigate the effect of longitudinal ventilation on the variation of burning rate in tunnel fires. The methanol square pool fires with heat release rate ranging from 3.57 kW to 10.95 kW were used. The burning rate of fuel was obtained by measured mass using load cell and temperature distribution were measured by K-type theomocouples in order to investigate smoke movement. The wind tunnel was connected with one side of the tested tunnel, and logitudinal ventilation velocity in the tested tunnel was controlled by power of the wind tunnel. In methanol fire case, the increase in ventilation velocity decreased the turning rate due to the direct cooling of fire plume. For the same dimensionless velocity(V), homing rate decreased as the size of pool fire increased.

• Journal of the Korean Society of Safety
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• v.37 no.4
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.8 no.4
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• pp.68-78
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• 1999

Heat transfer performance is studied for boiling and condensation of R-11 on integral-fin tubes. Nine tubes with trapezoidal integral-fins having fin densities from 748 to 1654fpm and 10,30 grooves and finned tubes with caves of 0.55 and 0.64 mm height respectively are tested. in case of condensation CFC-11 condensates at saturation stat of 32$^{\circ}C$ on the outside surface cooled by inside cooling water flows. And in case of boiling the refrigerant evaporates at a saturation state of 1 bar on the outside tube surface and heat is supplied by hot water which circulates inside of the tube,. The tube having fin transfer coefficient concerns fin tubes with caves show higher valve than low fin tube having find density of 1299fpm and 30grooves. The overall heat transfer coefficient of fin tube with caves is about 5155 W/mK at 2.8m/s of water velocity, The value is abuot 2.7 times higher than plain tube and 1.3 times higher than low fin tube having fin density of 1299fpm and 30 grooves.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.8 no.4
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• pp.65-65
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• 1999

Heat transfer performance is studied for boiling and condensation of R-11 on integral-fin tubes. Nine tubes with trapezoidal integral-fins having fin densities from 748 to 1654fpm and 10,30 grooves and finned tubes with caves of 0.55 and 0.64 mm height respectively are tested. in case of condensation CFC-11 condensates at saturation stat of 32℃ on the outside surface cooled by inside cooling water flows. And in case of boiling the refrigerant evaporates at a saturation state of 1 bar on the outside tube surface and heat is supplied by hot water which circulates inside of the tube,. The tube having fin transfer coefficient concerns fin tubes with caves show higher valve than low fin tube having find density of 1299fpm and 30grooves. The overall heat transfer coefficient of fin tube with caves is about 5155 W/mK at 2.8m/s of water velocity, The value is abuot 2.7 times higher than plain tube and 1.3 times higher than low fin tube having fin density of 1299fpm and 30 grooves.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.237-248
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• 2012

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

• Journal of computational fluids engineering
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• v.13 no.3
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• pp.21-27
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• 2008

The reactor head of a sodium-cooled fast reactor KALIMER-600 should be cooled during the reactor operation in order to maintain the integrity of sealing material and to prevent a creep fatigue. Analyzing turbulent natural convection flow in the cover gas region of reactor vessel with the commercial CFD code CFX10.0, the cooling requirement for the reactor head and the performance of the insulation plate were assessed. The results showed that the high temperature region around reactor vessel was caused by the convective heat transfer of Helium gas flow ascending the gap between the insulation plate and the reactor vessel inner wall. The insulation plate was shown to sufficiently block the radiative heat transfer from pool surface to reactor head to a satisfactory degree. More than $32.5m^3$/sec of cooling air flow rate was predicted to maintain the required temperature of reactor head.

• 한국전산유체공학회:학술대회논문집
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• 2004.03a
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• pp.208-213
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• 2004

The HANARO, multi-purpose research reactor, 30 MWth open-tank-in- pool type, is under normal operation since it reached the initial critical in February 1995. The HANARO is planning to produce a fission moly-99 of radio isotopes, a mother nuclide of Tc-99m, a medical isotope and is under developing a target handling tool for loading and unloading it in a circular flow tube (OR-5). A guide tube is extended from the reactor core to the top of the reactor chimney for easily un/loading a target under the reactor normal operation. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube by jet flow. This paper is described an analytical analysis to calculate the hole size of a orifice inserted in the circular irradiation hole and to study the flow characteristics through the guide tube under reactor normal operation and loading the target. As results, the results show that the hole size of orifice was 31 mm of the inner diameter to suppress the guide tube jet flow and the coolant safely cooled the target of fission moly after inserting the orifice to the flow tube.