• 한국가스학회지
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• 제7권4호
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• pp.14-19
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• 2003

일반적으로 발생하는 사고중의 가스폭발사고는 일정한 상태로의 혼합과정보다는 불균일하게 이루어진 상태로 폭발이 발생하고 있다. 본 연구에서는 불균일 상태에서의 불균일 농도에 의한 폭발현상을 실제 실내의 축소형으로 폭발통을 제작, 모형화하여 여러현상을 재현함으로서 실제 사고시의 폭발현상을 예측할 수 있었다. 그 결과, 어떤 공간에서 가스가 누출될 경우 내부의 가스의 혼합정도는 누출구의 크기 및 누출속도에 따라 영향을 받으며, 불균일 정도가 클수록 폭발압력 및 압력상승속도는 낮아지고 폭발압력은 감소하지만 화염의 체류시간이 증가하여 폭발화염의 복사열에 의한 폭발 후에는 화재의 위험성이 증가함을 알 수 있었다.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2003년도 하계학술대회 논문집 Vol.4 No.2
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• pp.1135-1138
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• 2003

The transformer is major equipment in power receiving and substation facilities. Necessary conditions required for the transformer are compactness, lightness, high reliability, economic advantages, and easy maintenance. The pole-mount transformer installed in distribution system is acting direct role in supply of electric power and it is electric power device should drive for long term. Most of modem transformer are oil-filled transformer and accident is happening considerable. The mold transformers have been widely used in underground substations in large building and have some advantages in comparison to oil-transformer, that is low fire risk, excellent environmental compatibility, compact size and high reliability. In addition, the application of mold transformer for outdoor is possible due to development of epoxy resin. The mold transformer generally has cooling duct between low voltage coil and high voltage coil. A mold transformer made by one body molding method has been developed for small size and low loss. One body molding transformer needs some cooling method because heat radiation between each winding is difficult. In this paper, The thermal analysis of pole mount mold transformer with one body molding by duct condition is investigated and the test result of temperature rise is compared with simulation data.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1845-1850
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• 2022

In-vessel corium retention (IVR) during external reactor vessel cooling (ERVC) is a key severe accident management strategy adopted in advanced nuclear power plants. The injection of nanofluids has been regarded as a means of enhancing CHF when using the IVR-ERVC strategy to safeguard high-power nuclear reactors. However, a critical practical concern is that various types of debris flowing from the contaminant sump during operation of an ERVC system might degrade CHF enhancement by nanofluids. Our objective here was to experimentally assess the viability of nanofluid use to enhance CHF in practical ERVC contexts (e.g., when fluids contain various types of debris). The types and characteristics of debris expected during IVR-ERVC were examined. We performed pool boiling CHF experiments using nanofluids containing these types of debris. Notably, we found that debris did not cause any degradation of the CHF enhancement characteristics of nanofluids. The nanoparticles are approximately 1000-fold smaller than the debris particles; the number of nanoparticles in the same volume fraction is 1 billion-fold greater. Nanofluids increase CHF via porous deposition of nanosized particles on the boiling surface; this is not hindered by extremely large debris particles.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1049-1078
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• 2021

The review summarizes the published data on the widely applied electron-beam, laser-beam, as well as resistance upset, projection, and spot welding of zirconium alloys for nuclear applications. It provides the results of their analysis to identify common patterns in this area. Great attention has been paid to the quality requirements, the edge preparation, up-to-date equipment, process parameters, as well as post-weld treatment and processing. Also, quality control and weld repair methods have been mentioned. Finally, conclusions have been drawn about a significant gap between the capabilities of advanced welding equipment to control the microstructure and, accordingly, the properties of welded joints of the zirconium alloys and existing algorithms that enable to realize them in the nuclear industry. Considering the ever-increasing demands on the high-burnup accident tolerant nuclear fuel assemblies, great efforts should be focused on the improving the welding procedures by implementing predefined heat input cycles. However, a lot of research is required, since the number of possible combinations of the zirconium alloys, designs and dimensions of the joints dramatically exceeds the quantity of published results on the effect of the welding parameters on the properties of the welds.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2836-2846
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• 2020

During severe nuclear power plant (NPP) accidents, a H₂/CO mixture can be generated in the reactor pressure vessel by core degradation and in the containment as well by molten corium-concrete interaction. In spite of its importance, a state-of-the-art methodology predicting H₂/CO combustion risk relies predominantly on empirical correlations. It is therefore necessary to develop a proper methodology for flammability evaluation of H₂/CO mixtures at ex-vessel phases characterized by three factors: CO concentration, high temperature, and diluents. The developed methodology adopted Le Chatelier's law and a calculated non-adiabatic flame temperature model. The methodology allows the consideration of the individual effect of the heat transfer characteristics of hydrogen and carbon monoxide on low flammability limit prediction. The accuracy of the developed model was verified using experimental data relevant to ex-vessel phase conditions. With the developed model, the prediction accuracy was improved substantially such that the maximum relative prediction error was approximately 25% while the existing methodology showed a 76% error. The developed methodology is expected to be applicable for flammability evaluation in chemical as well as NPP industries.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4003-4013
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• 2021

During the long-term cooling (LTC) phase of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), water is supplied from the containment sump to the reactor coolant system (RCS) by the flooded sump water to the Reactor Vessel (RV) through the broken pipes. As part of the technical efforts for resolving GSI-191 [( Reid and Crytzer, May. 2007) 1, consideration is needed for the consequences of debris penetrating the sump screen and propagating downstream into the RV. Injection of debris (fiberglass) into the RV during the LTC recirculation phase needs special attention to assure that reactor core cooling is maintained. The point of concern is the potential for debris to adversely affect the reactor core flow paths or heat transfer [2]. However, all the experiments for proving the coolability of RV have been done with the assumption of the most of debris would be transferred to the RV and the bottom nozzle of the FAs. The purpose of the tests is to quantify the amount of the debris that would be accumulated at the lower plenum and the debris that passes through the FAs since non-conservatism of other researches assumptions that have been used in the past experimental or analytical programs.

• International Journal of Advanced Culture Technology
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• 제11권3호
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• pp.346-357
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• 2023

The purpose of this study is to suggest improvement plans for prevention regulations by reflecting the toxicity, fire and explosion effects of hazardous materials factories and surrounding areas using an off-site consequence assessment program. Regarding the effects of the hydrogen cyanide leak accident, which is the 1st petroleum of the 4th class flammable liquid, Areal Locations of Hazardous Atmospheres (ALOHA) program was used to compare and analyze the extent of damage effects for toxicity, overpressure, and radiation. As a result, the toxicity was analyzed to exceed 5km in the area with Acute exposure guideline level (AEGL)-2 concentration or higher, the overpressure was 103m in the range of 1 psi or more, and the radiant heat was analyzed to be 724m in the range of 2kw/m² or more. Toxicity and radiation affected the area outside the hazardous material storage area, but the overpressure was limited to the inside of the hazardous material storage area. Therefore, we propose to improve the safety management of hazardous materials by conducting a risk assessment for hazardous materials and reflecting the results in internal and external emergency response plans to prepare prevention regulations.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2835-2843
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• 2023

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4159-4169
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• 2022

Hydrogen mitigation using Passive Autocatalytic Recombiners (PARs) has been widely accepted methodology inside reactor containment of accident struck Nuclear Power Plants. They reduce hydrogen concentration inside reactor containment by recombining it with oxygen from containment air on catalyst surfaces at ambient temperatures. Exothermic heat of reaction drives the product steam upwards, establishing natural convection around PAR, thus invoking homogenisation inside containment. CFD models resolving individual catalyst plate channels of PAR provide good insight about temperature and hydrogen recombination. But very thin catalyst plates compared to large dimensions of the enclosures involved result in intensive calculations. Hence, empirical correlations specific to PARs being modelled are often used in integral containment studies. In this work, an experimentally validated CFD model of PAR has been employed for developing an empirical correlation for Indian PAR. For this purpose, detailed parametric study involving different gas mixture variables at PAR inlet has been performed. For each case, respective values of gas mixture variables at recombiner outlet have been tabulated. The obtained data matrix has then been processed using regression analysis to obtain a set of correlations between inlet and outlet variables. The empirical correlation thus developed, can be easily plugged into commercially available CFD software.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4759-4775
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• 2022

The RVACS is a versatile and robust safety system driven by two natural circulations: in-vessel coolant and ex-vessel air. To observe interaction between the two natural circulations, SINCRO-IT facility was designed with two different similarity laws simultaneously. Bo' based similarity law was employed for the in-vessel, while Ishii's similarity law for the ex-vessel excluding the radiation. Compared to the prototype, the sodium and air system, SINCRO-IT was designed with Wood's metal and air, having 1:4 of the length reduction, and 1.68:1 of the time scale ratio. For the steady state, RV temperature limit was violated at 0.8% of the decay heat, while the sodium boiling was predicted at 1.3%. It showed good accordance with the system code, TRACE. For an arbitrary re-criticality scenario with RVACS solitary operation, sodium boiling was predicted at 25,100 s after power increase from 1.0 to 2.0%, while the system code showed 30,300. Maximum temperature discrepancy between the experiments and system code was 4.2%. The design and methodology were validated by the system code TRACE in terms of the convection, and simultaneously, the system code was validated against the simulating experiments SINCRO-IT. The validated RVACS model could be imported to further accident analysis.