• 한국안전학회지
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• 제20권2호
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2460-2470
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• 2020

To study the thermophysical and neutronic properties of thorium-plutonium fuel, a conceptual design of a hybrid facility consisting of a subcritical Th-Pu reactor core and a source of additional D-D neutrons that places on the axis of the core is proposed. The source of such neutrons is a column of high-temperature plasma held in a long magnetic trap for D-D fusionreactions. This article presents computer simulation results of generation of thermonuclear neutrons in the plasma, facility neutronic properties and the evolution of a fuel nuclide composition in the reactor core. Simulations were performed for an axis-symmetric radially profiled reactor core consisting of zones with various nuclear fuel composition. Such reactor core containing a continuously operating stationary D-D neutron source with a yield intensity of Y = 2 × 10¹⁶ neutrons per second can operate as a nuclear hybrid system at its effective coefficient of neutron multiplication 0.95-0.99. Options are proposed for optimizing plasma parameters to increase the neutron yield in order to compensate the effective multiplication factor decreasing and plant power in a long operating cycle (3000-day duration). The obtained simulation results demonstrate the possibility of organizing the stable operation of the proposed hybrid 'fusion-fission' facility.

• 한국소음진동공학회논문집
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• 제25권1호
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• pp.48-57
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• 2015

제 4세대 원자로로 개발되고 있는 소듐냉각고속로는 냉각재로 사용되는 소듐의 물과 공기에 의한 반응을 방지하기 위하여 원자로 헤드의 개방 없이 회전 구동만을 이용하여 핵연료 교환을 수행한다. 따라서 핵연료 교환을 위해서는 노심과 상부내부구조물 사이의 공간에 원자로 헤드의 회전 구동을 방해하는 장애물의 존재 여부를 확인하는 검사가 반드시 선행되어야 하는데, 소듐의 불투명성으로 인해 통상적인 광학 장비를 사용한 검사방법으로는 장애물의 확인이 어렵고, 초음파를 이용한 검사방법이 적용되어야 한다. 이 연구에서는 불투명한 소듐 중에서 노심과 상부내부구조물 사이의 장애물 존재 여부를 확인하기 위하여 판형 웨이브가이드 초음파 센서를 적용한 원격주사 검사기법을 개발하였다. 웨이브가이드 센서는 원자로헤드 상부의 저온 구간에 초음파 트랜스듀서를 설치하고 길이가 긴 웨이브가이드를 사용하여 노심 상부의 고온 고방사능 소듐 내부로 초음파를 전파시키기 때문에, 고온 고방사능에 의한 손상 없이 장시간 적용이 가능한 장점을 가지고 있다. 먼저, 수평 방향으로 초음파를 방사시킬 수 있는 10 m 길이의 수평 빔 웨이브가이드 센서를 설계 제작하였고, 제작된 센서의 특성 분석을 위한 빔 프로파일 측정 및 기초 실험을 수행하여 방사되는 초음파의 빔 폭과 전파 거리를 평가하였다. 또한, 실제 장애물로 작용할 수 있는 부품들의 형상인 원형 목표물에 대한 원격탐지 성능시험을 수행하여 개발된 원격주사 기술의 유용성을 평가하였다.

• 한국수소및신에너지학회논문집
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• 제33권4호
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• pp.419-429
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• 2022

Effects of operating variables on temperature profile and performance of 3 MWth chemical looping combustion system were estimated by mass and energy balance analysis based on configuration and dimension of the system determined by design tool. Air reactor gas velocity, fuel reactor gas velocity, solid circulation rate, and solid input percentage to fluidized bed heat exchanger were considered as representative operating variables. Overall heat output and oxygen concentration in the exhaust gas from the air reactor increased but temperature difference decreased as air reactor gas velocity increased. Overall heat output, required solid circulation rate, and temperature difference increased as fuel reactor gas velocity increased. However, overall heat output and temperature difference decreased as solid circulation rate increased. Temperature difference decreased as solid circulation rate through the fluidized bed heat exchanger increased. Effect of each variables on temperature profile and performance can be determined and these results will be helpful to determine operating range of each variable.

• 한국대기환경학회:학술대회논문집
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• 한국대기환경학회 2004년도 추계학술대회 논문집
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• pp.163-165
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• 2004

• 대한전기학회:학술대회논문집
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• 대한전기학회 2011년도 제42회 하계학술대회
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• pp.1814-1815
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• 2011

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2006년도 추계학술대회 논문집
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• pp.33-34
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• 2006

• 한국식품과학회지
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• 제31권3호
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• pp.593-599
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• 1999

Biogenic amine류는 어육의 부패과정 중 생성되는 성분으로 이를 간편하게 측정하기 위하여 diamine oxidase를 amino기가 관능기인 다공질의 Chitopearl beads에 고정화한 후 이를 충진한 효소 reactor를 제조하였다. 고정화 효소 reactor를 이용한 biogenic amine 연속계측 시스템을 사용하여 putrescine 기질에 대해 15.0 mM까지 검량곡선을 얻을 수 있었다. 이 때, biogenic amine 센서의 최대 감응도는 $35^{\circ}C$, 0.05 M 인산 완충용액(pH 7.2)을 사용한 경우 얻을 수 있었다. Biogenic amine 센서로 측정 시 방해물질로 작용할 수 있는 성분을 조사한 결과, ATP분해산물 및 TMA 등에는 거의 영향을 받지 않아 측정에 무리가 없음을 보여주었고, 0.8%까지의 식염에 대해서는 감응도가 19.2%이하로 감소하였으며 alanine과 glycine을 제외한 아미노산에 대하여 9.8% 이하의 감응도 상승을 나타내었다. 염류 및 아미노산에 의한 감응도 감소 및 증가 현상은 시료를 0.4 M perchloric acid에 의해 전처리하면 해결할 수 있었다. 여러 종류의 polyamines가 공존하는 경우 putrescine 기준으로 하여 putrescine 검량곡선의 26.7% 범위내에서 polyamines의 측정이 가능하였다. Diamine oxidase가 고정화된 Chitopearl 효소 reactor를 이용하는 경우 biogenic amines의 함량 유무를 30분 이내에 측정할 수 있었다.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2004년도 ICCAS
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• pp.1116-1120
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• 2004

In the nuclear power plant, there are several cylindrical vessels such as reactor vessel, pressuriser and so on. The vessels are usually constructed by welding large rolled plates, forged sections or nozzle pipes together. In order to assure the integrity of the vessel, these welds should be periodically inspected using sensors such as ultrasonic transducer or visual cameras. This inspection is usually conducted under water to minimize exposure to the radioactively contaminated vessel walls. The inspections have been performed by using a conventional inspection machine with a big structural sturdy column, however, it is so huge and heavy that maintenance and handling of the machine are extremely difficult. It requires much effort to transport the system to the site and also requires continuous use of the utility's polar crane to move the manipulator into the building and then onto the vessel. Setup beside the vessel requires a large volume of work preparation area and several shifts to complete. In order to resolve these problems, we have developed an underwater mobile robot guided by the laser pointer, and performed a series of experiments both in the mockup and in the real reactor vessel. This paper introduces our robotic inspection system and the laser guidance of the mobile robot as well as the results of the functional test.

• 한국안전학회지
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• 제32권2호
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• pp.147-152
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• 2017

The external reactor vessel cooling (ERVC) is well known strategy to mitigate a severe accident at which nuclear fuel inside the reactor vessel is molten. In order to compare the heat removal capacity of ERVC between the nuclear reactor designs quantitatively, numerical method is often used. However, the study for ERVC using computational fluid dynamics (CFD) is still quite scarce. As a validation study on the numerical prediction for ERVC using CFD, the subcooled boiling flow and natural circulation of coolant at the ULPU-V experiment was simulated. The commercially available CFD software ANSYS-CFX was used. Shear stress transport (SST) model and RPI model were used for turbulence closure and wall-boiling, respectively. The averaged flow velocities in the downcomer and the baffle entry under the reactor vessel lower plenum are in good agreement with the available experimental data and recent computational results. Steam generated from the heated wall condenses rapidly and coolant flows maintains single-phase flow until coolant boils again by flashing process due to the decrease of saturation temperature induced by higher elevation. Hence, the flow rate of coolant natural circulation does not vary significantly with the change of heat flux applied at the reactor vessel, which is also consistent with the previous literatures.