• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.82.2-82.2
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• 2003

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.185-194
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• 2006

Transient operations for an integral type reactor, SMART-P, have been experimentally investigated using a thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), in order to verify the system design and performance of the SMART-P, a pilot plant of SMART. The VISTA facility was subjected to various accident conditions such as feedwater increase and decrease, loss of coolant flow, and control rod withdrawal accidents in order to elucidate the thermal-hydraulic responses following such accidents and finally to verify the system design of the SMARTP. Full functional control logics have been implemented in the VISTA facility in order to control the required control action for an accident simulation. As one of the sensitivity tests to verify the PRHRS performance, the effects of the initial water level in the compensation tank are experimentally investigated. When the initial water level is 16%, the water is quickly drained and nitrogen gas is then introduced into the PRHR system, resulting in deterioration of the PRHRS performance. It is thus found that nitrogen ingression should be prevented to ensure stable PRHRS operation.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.2053-2058
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• 2004

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.148-148
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.629-630
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• 2005

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.781-782
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• 2005

• Nuclear Engineering and Technology
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• 제41권1호
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• pp.53-62
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• 2009

A set of experiments has been conducted on the performance sensitivity of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART, by using a high temperature and high pressure thermal-hydraulic test facility, the VISTA facility. In this paper the effects of the opening delay of the PRHRS bypass valves and the closing delay of the secondary system isolation valves, and the initial water level and the initial pressure of the compensating tank (CT) are investigated. During the reference test a stable flow occurs in a natural circulation loop that is composed of a steam generator secondary side, a secondary system, and a PRHRS; this is ascertained by a repetition test. When the PRHRS bypass valves are operated 10 seconds later than the secondary system isolation valves, the primary system is not properly cooled. When the secondary system isolation valves are operated 10 or 30 seconds later than the PRHRS bypass valves, the primary system is effectively cooled but the inventory of the PRHRS CT is drained earlier. As the initial water level of the CT is lowered to 16% of the full water level, the water is quickly drained and then nitrogen gas is introduced into the PRHRS, resulting in the deterioration of the PRHRS performance. When the initial pressure of the PRHRS is at 0.1MPa, the natural circulation is not performed properly. When the initial pressures of the PRHRS are 2.5 or 3.5 MPa, they show better performance than did the reference test.

• 헤리티지:역사와 과학
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• 제55권1호
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• pp.5-34
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• 2022

본 연구는 청대 황가능침원 12기를 대상으로 입지와 풍수 그리고 형태와 구조 등이 어우러진 시설 배치상의 특징을 고찰함으로써 청대 황가능침원 고유의 조영 경관적 특성을 구명하고자 한 것이다. 문헌조사와 현장 관찰 및 인터뷰, 영상 및 도면 비교 그리고 귀납 대비적 분석을 통해 청나라 시대 황가능침원의 조영 경관 특성을 자연환경, 입지 및 풍수 형세, 조영 형태와 구조로 구분하여 분석하였다. 그 결과 풍수 형국, 중축선에 따른 축선 연장거리, 능침 폭, 기점과 종점의 고저의 차에 따른 시설배치의 정형성, 능침의 좌향, 진입공간-제향공간-능침공간의 공간 비율 등의 분석에 따라 청대 황가능침원의 입지형국과 배치 계획상의 특성을 도출하였다. 또한 능침 시설물의 종류와 배치 순서 그리고 석상생의 분포 및 수량 및 종류 분석을 통해 능침의 시설배치적 특성을 도출하는 한편 명나라 시대와 다른 몇 가지 조영 특성을 밝힐 수 있었다. 또한 수계를 통한 공간분할 효과, 중축선에서의 광경(框景)과 협경(夾景)의 조망 효과가 소밀대비(疏密對比)와 함께 청대 황가능침원에서 발견되는 대표적 조망경관 효과임을 확인하였다.