• Journal of Energy Engineering
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• v.10 no.4
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• pp.342-348
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• 2001

A new design concept for a decay heat removal system in a liquid metal reactor is proposed. The new design utilizes a thermosyphon to enhance the heat removal capacity and its heat transfer characteristics are analyzed against the current PSDRS (Passive Safety Decay heat Removal System) in the KAL IMER (Korea Advanced LIquid MEtal Reactor) design. The preliminary analysis results show that the new design with a thermosyphon yields substantial increase of 20∼40% in the decay heat removal capacity compared to the current design that do not have the thermosyphon. The new design reduces the temperature rise in the cooling air of the system and helps the surrounding structure in maintaining its mechanical integrity for long term operation at an accident. Also the analysis revealed the characteristics of the interactions among various heat transfer modes in the new design.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.143-148
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• 2007

Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A dry storage cask to safely store the spent nuclear fuel should be able to adequately emit the decay heat from the spent nuclear fuel. Therefore, heat removal tests using a half scale dry storage cask have been performed to estimate the heat transfer characteristics of a dry storage cask under normal, off-normal, and accident conditions. In the normal condition, the heat transfer rate to an ambient atmosphere by convective air through a passive heat removal system reached 83%. Accordingly, the passive heat removal system is designed well and works adequately. In the off-normal condition, the influence of a half blockage in the inlet on the temperature appears minimal. In the accident condition, the temperature rose for 12 hours after the accident, but the temperature rise steadied after 36 hours.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.734-741
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• 2020

A new air-cooled waste heat removal system with a direct contact heat exchanger was designed for SMRs requiring 200 MW of waste heat removal. Conventional air-cooled systems use fin structure causing high thermal resistance; therefore, a large cooling tower is required. The new design replaces the fin structure with a vertical string type direct contact heat exchanger which has the most effective performance among tested heat exchangers in a previous study. The design results showed that the new system requires a cooling tower 50% smaller than that of the conventional system. However, droplet formation on a falling film along a string caused by Rayleigh-Plateau instability decreases heat removal performance of the new system. Analysis of Rayleigh-Plateau instability considering drag force on the falling film surface was developed. The analysis results showed that the instability can be prevented by providing thick string. The instability is prevented when the string radius exceeds the capillary length of liquid by a factor of 0.257 under stagnant air and 0.260 under 5 m/s air velocity.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.395-409
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• 2000

Parametric studies were performed to assess the sensitivity in determining the maximum in-vessel heat removal capability from the core material relocated into the lower plenum of the reactor pressure vessel (RPV）during a core melt accident. A fraction of the sensible heat can be removed during the molten jet delivery from the core to the lower plenum, while the remaining sensible heat and the decay heat can be transported by rather complex mechanisms of the counter-current flow limitation (CCFL) and the critical heat flux (CHF）through the irregular, hemispherical gap that may be formed between the freezing oxidic debris and the overheated metallic RPV wall. It is shown that under the pressurized condition of 10MPa with the sensible heat loss being 50% for the reactors considered in this study, i.e. TMI-2, KORI-2 like, YGN-3&4 like and KNGR like reactors, the heat removal through the gap cooling mechanism was capable of ensuring the RPV integrity as much as 30% to 40% of the total core mass was relocated to the lower plenum. The sensitivity analysis indicated that the cooling rate of debris coupled with the sensible heat loss was a significant factor The newly proposed heat removal capability map (HRCM) clearly displays the critical factors in estimating the maximum heat removal from the debris in the lower plenum. This map can be used as a first-principle engineering tool to assess the RPV thermal integrity during a core melt accident. The predictive model also provided ith a reasonable explanation for the non-failure of the test vessel in the LAVA experiments performed at the Korea Atomic Energy Research Institute (KAERI), which apparently indicated a cooling effect of water ingression through the debris-to-vessel gap and the intra-debris pores and crevices.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3879-3891
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• 2021

A heat pipe residual heat removal system is proposed to be incorporated into the reactor driven subcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing and demonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of the system operation after the shutdown of the facility. One of the main components of the system is an air-cooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet the operation temperature and residual heat removal requirement of the facility. The heat pipe model developed in the previous work was adopted to simulate the designed heat pipe and assess the heat transport capability. 3D numerical simulation of the subcritical facility active zone was performed by the commercial CFD software STAR CCM + to investigate the operation characteristics of this proposed system. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident were simulated and analyzed. The results show that the residual heat removal system can provide sufficient cooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under the recommended operation temperature range and the heat flux is below all thermal limits. The facility peak temperature is also lower than the safety limits.

• Journal of environmental and Sanitary engineering
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• v.15 no.3
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• pp.37-43
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• 2000

The purpose of this experimental study examine characteristics of blood pigments removal of butchery wastewater by heat processed eggshell, compare activated carbon with its efficiency. Calcined eggshell were classified into four kinds of mesh as HPES-32(Heat Processed Eggshell 32 $mesh=500{\mu}m$), HPES-48($300{\mu}m$), HPES-150($180{\mu}m$) and HPES-150($106{\mu}m$). And two contacting process of CMFA(Complete Mixing Float Adsorption) and FLFA(Fixing layer Flow Adsorption) Were used for getting removal efficiency of blood pigments. In case of using CMFA process, the removal efficiency of blood pigments was occurred as HPES-80>HPES-150>HPES-32, but in case of using FLFA process was occurred as HPES-150>HPES-80>HPES-48>HPES-32. The two results between CMFA and FLFA were differ in strength of removal efficiency of blood pigments.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2088-2095
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• 2023

The safety of nuclear installations is largely determined by the tightness of fuel elements cladding. As the Fukushima nuclear accident showed, the main task in case of loss of power supply is to ensure reliable removal of residual heat release from spent fuel pool (SFP) with irradiated fuel assemblies (IFAs). The paper presents the results of calculated-experimental studies and thermal-hydraulic modeling of temperature storage modes of IFAs in SFP. Experimental studies of SFP's temperature regime and calculated evaluation of residual heat removal due to the thermal conductivity of building structures surrounding the SFP were performed. To ensure the safe operation of research reactors, it's necessary to know the IFA's residual heat power (RHP) in the reactor and SFP, which is determined depending on the operating time of fuel assemblies (FAs) and the IFAs calculated holding time. The FAs operating time depends on the reactor energy output. The IFAs calculated holding time is determined by the fuel burnup, U-235 mass in the fuel, and reactor utilization factor. The IFAs fuel burnup was calculated using the MCU-PTR program. Also presented are the RHP's calculation results using some of the empirical dependencies. The concept of a passive heat removal system (PHRS) based on thermosyphon's operating principle was proposed.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.260-266
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• 2015

Twenty sodium-cooled fast reactors (SFRs) have provided valuable experience in design, licensing, and operation. This paper summarizes the important safety criteria and safety guidelines of intermediate sodium systems, steam generators, decay heat removal systems and associated construction materials and in-service inspection. The safety criteria and guidelines provide a sufficient framework for design and licensing, in particular by new entrants in SFRs.

• KIEAE Journal
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• v.14 no.6
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• pp.81-86
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• 2014

In recent years, many researchers have considered the building energy consumption reduction accordingly to deal with abnormal climate changes and greenhouse gas reduction. However, the lighting energy use ratio has increased in spite of the development of the high efficiency lighting device. Therefore, the study aims to produce the LED lighting applications for the effective lighting heat removal by using the heat characteristics of LED lighting and analyzing the heat removal effect. In order to increase radiant heat efficiency, the heat pipe and heat sink was attached on PCB as LED lighting applications. Experiment was conducted to verify the temperature and air velocity of inside duct: thermocouples, anemometer. The heat removal effect of LED lighting applications was measured by observing the temperature of the lighting applications and the change of air velocity in duct. The experiment shows that the temperature change in the duct according to air velocity was $0.9{\sim}5.8^{\circ}C$. It is also concluded that heat removal was calculated from 33 to 81W.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.514-519
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• 1997

The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control roe drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation.