• 수도
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• 제23권5호통권80호
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• pp.13-22
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• 1996

• 대한기계학회논문집A
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• 제35권11호
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• pp.1499-1506
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• 2011

초고온가스로로부터 생성된 $950^{\circ}C$ 정도의 초고온 열을 이용하여 수소를 경제적이며 또한 대량으로 생산하려는 원자력수소생산시스템에서 중간열교환기는 원자로에서 생산된 초고온 열을 수소생산 공장으로 전달하는 핵심 기기중의 하나이다. 한국원자력연구원에서는 초고온가스로에 사용될 핵심 기기에 대한 성능시험을 위해 소형가스루프를 구축하였고 중간열교환기의 유력한 형태로 고려되고 있는 인쇄기판형 열교환기의 소형 시제품을 제작하였다. 본 연구는 인쇄기판형 열교환기 소형 시제품을 소형가스루프에서 시험하기 전에 루프 시험조건하에서 인쇄기판형 열교환기 소형 시제품의 고온 구조건전성을 미리 평가하기 위한 작업의 일환으로 수행한 결과, 즉 고온 구조해석 모델링, 거시적 열 해석 및 구조 해석 결과 등을 정리한 것이다. 해석 결과는 인쇄기판형 열교환기 소형 시제품 성능시험결과외 비교하고 향후 제작될 중형 시제품 설계/제작에 반영할 것이다.

• 한국압력기기공학회 논문집
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• 제8권1호
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• pp.33-38
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to a chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute has established a small-scale gas loop for the performance test on VHTR components and recently has manufactured a medium-scale PHE prototype made of Hastelloy-X. A performance test on the PHE prototype is scheduled in the gas loop. In this study, high-temperature structural analysis modeling, and macroscopic thermal and structural analysis of the medium-scale PHE prototype by imposing the established displacement boundary constraints in the previous research were carried out under the gas loop test condition. The results obtained in this study will be compared with performance test results.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.1-6
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature gas cooled Reactor) to the chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. Previous research on the high-temperature structural analysis of the small-scale PHE prototype had been performed only using parent material properties. In this study, high-temperature structural analysis using weld properties in weld zone was performed and the analysis results compared with the previous research.

• International Journal of Air-Conditioning and Refrigeration
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• 제9권2호
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• pp.77-84
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• 2001

Experimental and computational studies were carried out to investigate the turbulent heat transfer enhancement of the cooling system in nuclear reactor by large scale vortex generation. The large scale vortex motion was generated by rearranging the inclination angels of mixing vanes to the coordinate direction. Axial development of mean and turbulent velocities in the subchannels were measured by the 2-color LDV system. Eddy diffusivity concept based on $\kappa{-}\varepsilon$ model was employed to calculate the turbulent heat and momentum transfers in the subchannel. The turbulences generated by split mixing vanes has small length scales so that they maintain only about $10D_H$ after the spacer grid. On the other hand, the turbulences generated by the large scale vortex motions continue longer and remain up to $25D_H$ after the spacer grid.

• 설비공학논문집
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• 제12권11호
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• pp.1004-1011
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• 2000

Experimental and computational studies were carried out to confirm the turbulent enhancement of the cooling system of nuclear reactor by large scale vortex generation in nuclear fuel bundle. The large scale vortex motions were generated by rearranging the inclination angles of mixing vanes to the coordinate directions. Axial development of mean and turbulent velocities in the subchannels were measured by the 2-color LDV system. Eddy diffusivity heat flux model and $k-varepsilon$ model were employed to analyze the turbulent heat and fluid flows in the subchannel. The turbulence generated by split mixing vanes has small length scales so that they maintain only about $10 D_H$ after the spacer grid. On the other hand, the turbulences generated by the large scale vortex continue more and remain up to $25 D_H$after the spacer gird.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2001년도 춘계 학술발표회 논문집
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• pp.111-118
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• 2001

• 에너지공학
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• 제14권4호
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• pp.268-276
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• 2005

액체연료(중유)공급량 기준 20kg/hr규모의 반응로에서 증기분무 내부 혼합식 노즐을 이용하여 잔사유의 연소실험을 수행하였다. 본 실험에서 사용한 감압 잔사유는 점도가 높고 황함량, 잔류탄소와 금속성분의 함량도 높았다. 잔사유의 착화를 위해서는 반응로를 일정온도까지 예열하여야 했으며 이는 LPG를 이용하였다. 잔사유 공급량을 변화시키면서 축방향 및 반경방향의 로내 가스 온도, 주요 가스농도 및 채집된 고체 입자를 분석하였다. 잔사유의 주반응영역은 버너 팁으로부터 약 1 m 근방에서 형성되었으며 이는 축방향 가스 온도, 농도 분포 및 입자의 크기로부터 확인할 수 있었고, 반응로의 하류에서는 완전 확립된 온도분포를 보여주고 있었다. 고체 입자의 SEM 분석으로부터 잔류 탄소입자는 기공이 많은 형태를 띠고 있었다.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3580-3596
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• 2021

An integral test facility, PILLAR, was commissioned, aiming to provide valuable experimental results which can be referenced by system and component designers and used for the performance demonstration of liquid-metal-cooled, passive small modular reactors (SMRs) toward their licensing. The setup was conceptualized by a scaling analysis which allows the vertical arrangements to be conserved from its prototypic reactor, scaled uniformly in the radial direction achieving a flow area reduction of 1/200. Its final design includes several heater rods which simulate the reactor core, and a single heat exchanger representing the steam generators in the prototype. The system behaviors were characterized by its data acquisition system implementing various instruments. In this paper, we present not only a detailed description of the facility components, but also selected experimental results of both steady-state and transient cases. The obtained steady-state test results were utilized for the benchmark of a system code, achieving a capability of accurate simulations with ±3% of maximum deviations. It was followed by qualitative comparisons on the transient test results which indicate that the integral system behaviors in passive LBE-cooled systems are able to be predicted by the code.