• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1999년도 학회창립 10주년 기념 1999년도 가을 학술발표회 논문집
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• pp.869-881
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• 1999

Inchon International Airport Elevated Road Way is located between the Passenger Terminal Building and Transportaion Center which are Inchon International Airport core construction projects. The deck of the bridge is consists of 5-span or 6-span continuous pre-stressed concrete slab. Steel form has been used to enhance the quality of texture on concrete slab. Steel form has been used to enhance the quality of texture on concrete surface, lower surface of deck slab with the two way arch has been manufactured by highly professional manner in order to get an beautiful exterior architectural looks. The prestressed concrete deck slab is mass concrete structures with a high-specified concrete strength and a varying section in the range of 0.95-2.8m thickness. Therefore high risks of thermal cracking occurrence by heat of hydration highly are expected. To resolve such problem, we adopted type 1 cement and pipe cooking method at construction site through mass concrete specimen test and 3-dimensional analysis. For Pipe cooling we used 25mm diameter stainless pipes with wrinkles. Cooling pipe with spacing 50-60cm has been installed. And continuous pipe cooling with cooling water of 15$^{\circ}C$ was conducted for 2days. In present 8 span of all 29 spans construction has been completed. No thermal cracking heat hydration has been observed yet.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 Proceedings of the 4th Korea-China Joint Workshop on Nuclear Waste Management
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• pp.150-164
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• 2004

According to the Korean long-term plan for nuclear technology development, KAERI is conducting a few R&D projects related to the proliferation-resistant back-end fuel cycle. The R&D activities for the back-end fuel cycle are reviewed in this work, especially focusing on the study of the partitioning and transmutation(P&T) of long-lived radionuclides. The P&T study is currently being carried out in order to develop key technologies in the areas of partitioning and transmutation. The partitioning study is based on the development of pyroprocessing such as electrorefining and electrowinning because they can be adopted as proliferation-resistant technologies in the fuel cycle. In this study, various behaviors of the electrodeposition of uranium and rare earth elements in the LiCl-KCl electrorefining system have been examined through fundamental experimental work. As for the transmutation system, KAERI is studying the HYPER (HYbrid Power Extraction Reactor), a kind of subcritical reactor which will be connected with a proton accelerator. Up to now, a conceptual study has been carried out for the major elemental systems of the subcritical reactor such as core, transuranic fuel, long-lived fission product target, and the Pb-Bi cooling system, etc. In order to enhance the transmutation efficiency of the transuranic elements as well as to strengthen the reactor safety, the reactor core was optimized by determining its most suitable subcriticality, the ratio of height/diameter, and by introducing the concepts of optimum core configuration with a transuranic enrichment as well as a scattered reloading of the fuel assemblies.

• 대한기계학회논문집 C: 기술과 교육
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• 제3권3호
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• pp.209-215
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• 2015

노내계측계통의 설치 위치 및 케이블의 관통위치가 중대사고 대처계통에 미치는 영향을 노내 노심용융물 억류 및 원자로용기 외벽냉각 전략과 노외 노심용융물 냉각계통을 중심으로 조사하였다. 기존에 국내원전에서 주로 사용되었던 노내계측계통의 원자로 용기 하부탑재 및 ICI케이블의 원자로 용기하부 관통이 중대사고에 미치는 영향을 정리하고, 이러한 단점을 개선하기 위해 노내계측계통의 ICI 케이블이 원자로 용기 상부를 관통하는 상부탑재 노내계측계통의 장점을 기술하였다.

• 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.

• 에너지공학
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• 제25권2호
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• pp.45-51
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• 2016

연구용원자로에서 여러 수조 및 일차냉각계통 내부에 존재하는 냉각재를 정화시키기 위해 설치되는 수조수관리계통은 일차냉각계통 펌프가 정지한 후 원자로에서 발생하는 노심 붕괴열을 제거한다. 또한, 작업수조 내의 조사물과 사용후핵연료저장조 내에 저장된 사용후핵연료에서 발생하는 열을 제거하여 수조수의 온도를 제한 값 이내로 유지하는 기능도 수행한다. 본 연구에서는 수조수관리계통의 설계와 운전 방법을 설계 초기단계에서 결정하기 위해서 상용프로그램인 Flowmaster를 이용한 전산해석방법으로 수조수관리계통의 열교환기를 설계하고, 각 수조수의 온도를 시간에 따라 예측하였다.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3085-3094
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• 2022

The design concept of a Small High-flux Multipurpose LBE(Lead Bismuth Eutectic) cooled Fast Reactor (SHMLFR) was proposed in the paper. The primary cooling system of the reactor is forced circulation, and the fuel element form is arc-plate loaded high enrichment MOX fuel. The core is cylindrical with a flux trap set in the center of the core, which can be used as an irradiation channel. According to the requirements of the core physical design, a series of physical design criteria and constraints were given, and the steady and transient parameters of the reactor were calculated and analyzed. Regarding the thermal and hydraulic phenomena of the reactor, a simplified model was used to conduct a preliminary analysis of the fuel plates at special positions, and the temperature field distribution of the fuel plate with the highest power density under different coolant flow rates was simulated. The results show that the various parameters of SHMLFR meet the requirements and design criteria of the physical design of the core and the thermal design of the reactor. This implies that the conceptual design of SHMLFR is feasible.

• 한국정밀공학회지
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• 제29권11호
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• pp.1190-1198
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• 2012

A hot forming of large thick Al plate using a grid-type hybrid die is a process to make a shell plate for the production of a spherical LNG tank. This process is characterized by using a grid-typed die with an additional air cooling system for reducing the cooling time of the heated plate after hot forming. The process consists of the plate's feeding, heating, forming and cooling in detail and each of them is continuously performed along the rail. This paper was designed to propose the analytical and experimental methods for determining the convection and interfacial heat transfer coefficients required in hot forming analysis of Al plate. These values in the analysis are to reproduce numerically the cooling performance of grid-typed die and cooling device. Interfacial heat transfer was obtained from the heat transfer experiments for different pressures and inverse analysis method. To verify the efficiency of the coefficient values obtained from above methods, FE analysis and experiment of the hot spherical-forming process were conducted for a small-scaled model. The convection coefficient was also calculated from flow analysis of air released by cooling device within grid-typed die using ANSYS-CFX.

• 한국반도체및디스플레이장비학회:학술대회논문집
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• 한국반도체및디스플레이장비학회 2006년도 추계학술대회 발표 논문집
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• pp.132-137
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• 2006

With the help of the development of digital-multimedia in the middle of 1990's, FDP(Flat Panel Display) had attracted considerable attention. Collimation proximity exposure system that transfers the pattern on wafer or glass exactly using mask and light with appropriate wavelength is core process in semiconductor and liquid display element. The performances of resolution required in precision exposure system are evaluated by resolving power, depth of focus and storage area. Most of development has targeted on these three factors. The optical design including lamp house has played an important role on the performance of exposure process. In this study, we evaluate the cooling system, concerning on exposure device with mercury lamp among the kernel equipment for the production of LCD, to prevent the instability of lighting due to long term accumulation of excessive heating inside the lamp house. Numerical analysis is conducted on full-scale model. The characteristics of three-dimensional flow, pressure and temperature distribution on exposure system are graphically depicted to estimate the whole cooling system for lamp house and to establish the design criteria.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.