• Nuclear Engineering and Technology
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• 제42권6호
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• pp.656-661
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• 2010

Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER$^2$) facility, a thermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advanced nuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized with available pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separated by complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the rest of the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is found that the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accounted for; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptions used for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD also requires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience with liquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained and compared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducer located between the core inlet and outlet at a flow rate of 13.57kg/s.

• 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.

• 한국유체기계학회 논문집
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• 제10권1호
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• pp.7-13
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• 2007

This study proposes a modified bifurcation model with a computational fluid analysis according to variation of a bifurcation geometry. FLUENT is used for a calculation of the head losses in case of a generation and a pumping. The pressure, velocity field and turbulent intensity are simulated in a bifurcation. With consideration about these flow properties, we propose the modified model to improve a flow efficiency and reduce a sound. The proposed model is able to cut down a head loss by 45% when a generation and 36% when a pumping.

• 설비공학논문집
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• 제22권11호
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• pp.751-759
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• 2010

A newly developed type PCHE(Printed Circuit Heat Exchanger), which has a longitudinal corrugation flow channel, was fabricated using etching and diffusion bonding to evaluate its hydraulic performance. The pressure drop characteristics obtained from the experimental results are presented and the local friction factors associated with different hydraulic diameters and inclination angles are discussed. The results of a three-dimensional numerical simulation are presented, conducted using commercial CFD(Computational Fluid Dynamics) software at lower Reynolds number range. The numerical results were validated by experimental data obtained from helium gas experimental apparatus. The results of CFD prediction show fairly good agreement with the experimental data.

• 동력기계공학회지
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• 제18권4호
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• pp.11-16
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• 2014

Although smoke emitted from ships is not included in IMO(International Maritime Organiztion) regulation yet, it is one of the substance what is polluting mainly the air. Especially, its concentration is very high when an engine is started and a load is rapidly changed. This is caused by unburned fuel what is injected more than necessary quantity after combustion period. It is possible to decrease smoke concentration emitted at starting engine by controlling fuel injection quantity, but it is concerned that time to rated speed must be spent. Then a governor what can reduce the smoke concentration without a loss of time to rated speed is needed. We adopted a electro-hydraulic governor what can control dual fuel start limit function and achieved very low level of smoke concentration without greater the loss of time to rated speed.

• 한국농공학회지
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• 제14권1호
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• pp.2557-2569
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• 1972

Hydraulic computations to determine the elevation of canal bottom and mater surface for box type concrete culverts are discussed. Velocity and cross sectional area of flow are computed from Manning's formula. Aad then head loss and velocity head are considered to determine the elevation of bottom and water surface. For stress analysis, 13.5 ton live load and earth pressure are considered. Also longitudinal stress of box culverts is checked.

• 한국관개배수논문집
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• 제3권2호
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• pp.44-52
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• 1996

To make a real hydraulic simulation model, Hazen-Williams' coefficient must be calibrated using the measured discharges and hydraulic head loss. In this paper, calibrating method of Hazen-Williams' coefficient was proposed for pipeline system. The calibra

• 한국산학기술학회논문지
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• 제14권11호
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• pp.5956-5963
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• 2013

친환경 하천사업의 활성화로 식생매트의 사용은 증가하고 있으나 수리적 안정성에 대한 평가기법은 제시되지 않은 실정이다. 본 연구는 호안용 식생매트 제품의 객관적인 수리안정성 시험 기법 개발을 위해 수행하였다. 이를 위해서 식생매트 2종에 대한 실규모 실험을 수행하여 수리량을 측정하고 분석하여 작용 소류력을 계산하였다. 작용 소류력에 대한 토양손실평가를 위하여 지상라이다에 의한 측량을 수행하고 실험 전후의 하상고 변화를 평가하여 토양손실지수(CSLI)를 산정하였으며, 작용 소류력과 함께 도시하여 허용 소류력을 정량적으로 평가하였다. 하상고에 대한 정밀 측량 결과 분석에 의해서 식생매트가 안정한 경우에는 하상 변동이 국부적으로 제한되나, 불안정으로 판정되는 경우에는 식생매트 하부에서 비교적 큰 규모의 하상 변동이 발생하며 이는 자연 하상의 거동과 유사함을 확인하였다. 이상의 연구를 통해서 ASTM D 6040에 의한 식생매트의 허용 소류력 평가가 파괴 메카니즘 및 토양손실판정 기준에서 유효함을 확인할 수 있었다.

• 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.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2002-2010
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• 2024

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.