• 한국초전도ㆍ저온공학회논문지
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• 제10권4호
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• pp.43-45
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• 2008

The thermal characteristics of the helium circulation by a cryocooler are presented. This study is motivated mainly by our recent development of a closed-loop cooling system for Cyclotron K120 superconducting magnets without any replenishment of the cryogen. A channel is attached on the outer surface of the magnet form and the liquid helium passes through inside of the channel in order to cool the super conducting coils indirectly. A two-stage cryocooler as a heat sink is located at the top to recondense helium coming from the superconducting magnet form. The heat transfer in the natural circulation loop is discussed and the main dimensions of cooling system are determined.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.330-338
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• 2023

In this study, impacts of an enhanced-moderation Fuel Assembly (FA) named Truly Optimized PWR (TOP) lattice, which is modified based on the standard 17 × 17 PWR FA, are investigated in a natural circulation Soluble-Boron-Free (SBF) Small Modular Reactor (SMR). Two different TOP lattice designs are considered for the analysis; one is with 1.26 cm pin pitch and 0.38 cm fuel pellet radius, and the other is with 1.40 cm pin pitch and 0.41 cm fuel pellet radius. The NuScale core design is utilized as the base model and assumed to be successfully converted to an SBF core. The analysis is performed following the primary coolant circulation loop, and the reactor is modelled as a single channel for thermal-hydraulic analyses. It is assumed that the ratio of the core pressure drop to the total system pressure drop is around 0.3. The results showed that the reactor power could be increased by 2.5% and 9.8% utilizing 1.26/0.38 cm and 1.40/0.41 cm TOP designs, respectively, under the identical coolant inlet and outlet temperatures as the constraints.

• 대한기계학회논문집B
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• 제22권8호
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• pp.1083-1090
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• 1998

A kind of three-way check valve, so called hydraulic calve was proposed for the substitute of the density lock of passive reactor such as SPWR (System-Integrated Pressurized Water Reactor). The function of the valve are the separation of the borated water from main coolant loop for normal operation and the insurge of the water into the loop for shutdown and the removal of the decay power when the main coolant flow rate is not enough. To verify the operability and the characteristics of the valve, experimental works were executed with 1/3 scale model calve. Also a diffuser model was proposed for the theoretical analysis of the valve.

• 대한기계학회논문집
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• 제8권5호
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• pp.416-424
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• 1984

본 연구에서는 다회로의 단상(single-phase) 자연순환계에 관한 실험 및 수치 해석적 연구로서, RWR에서의 자연순환 현상을 모사할 수 있는 fast-running code를 개 발하고 이를 실험을 통하여 입증하며, 또한 자연순환계에서 일어나는 여러 현상을 정 성적으로 관찰하는 것을 목적으로 삼았다. 이론부분은 저자의 다른 논문에 발표되었 으므로 여기에서는 요약하여 소개하며, 실험은 2-회로 PWR(고리 1호기) 1차계통을 약 1/15로 축소시킨 실험장치에서 행하였다.

• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1617-1628
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• 2017

In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

• 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.

• Nuclear Engineering and Technology
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• 제27권4호
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• pp.483-490
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• 1995

PKL III A 4.4 실험은 "증기발생기 세관파열사고 후 소외전원 가용의 조건 하에서 발전소가 비상냉각수 주입없이 수작동에 의해 제어될 수 있음을 확인하는 것이다. 실험 모사에 따른 NLOOP Code의 제한이나 능력의 검증을 위해, 실험에서 얻어진 PKL 설비의 거동은 NLOOP의 결과와 상호 비교되었다. NLOOP 코드는 한국원자력연구소와 독일 SIEMENS/KWU사에 의해 Westinghouse 형 발전소의 과도현상 해석용으로 개발되었으며, PKL III 설비모사를 위해 적절히 수정되었다. 자연대류에 의한RCS Loop의 냉각수 유량과 격리된 RCS Loop에서의 자연대류 중단현상을 특별히 주의깊게 연구하였다. 실험과 계산 결과의 비교는 NLOOP 코드의 의사능가 문제점들을 보여준다.보여준다.

• 대한기계학회논문집B
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• 제33권4호
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• pp.235-241
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• 2009

Experimental studies on single-phase toroidal circulation loop(thermosyphon) have been performed in the present study with Ag-nanofluids as a working fluids. The present paper deals with an experimental study on the heat transfer behavior of single-phase toroidal loop. Toroidal loop charged with nanofluid has been constructed and a number of tests have been carried out. Different geometric parameter, e.g., orientation has been investigated. The tests were conducted employing two fluids: distilled water and Ag-nanofluid of various volume concentrations. The experiments at Rayleigh number from $10^5$ to $10^6$ showed a systematic and slight deterioration in natural convective heat transfer. It was observed that the deterioration due to the particle concentration was in the range of 5-10%. At a given particle concentration of 0.05%, abrupt decrease in the Nusselt number and the Raleigh number was observed. The present study with toroidal loop shows that the application of nanofluids for heat transfer intensification should not be decided only by the effective thermal conductivity with increasing particle concentration.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.889-896
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• 2020

The vertically inverted U-tube steam generator (UTSG) is widely used in the pressurized water reactor (PWR). The reverse flow behavior generally exists in some U-tubes of a steam generator (SG) under both single- and two-phase natural circulations (NCs). The behavior increases the flow resistance in the primary loop and reduces the heat transfer in the SG. As a consequence, the NC ability as well as the inherent safety of nuclear reactors is faced with severe challenges. The theoretical models for calculating single- and two-phase flow pressure drops in U-tubes are developed and validated in this paper. The two-phase reverse flow characteristics in two types of SGs are investigated base on the theoretical models, and the effects of the U-tube height, bending radius, inlet steam quality and primary side pressure on the behavior are analyzed. The conclusions may provide some promising references for SG optimization to reduce the disadvantageous behavior. It is also of significance to improve the NC ability and ensure the PWR safety during some accidents.