• 대한기계학회논문집A
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• 제35권11호
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• pp.1383-1389
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• 2011

제로터 펌프는 용적식 펌프로서 유압 및 수압장치에 적합하며, 다양한 분야에서 널리 사용되어지고 있다. 이러한 이유로 많은 연구자들에 의해 연구되어지고 있다. 본 연구에서는 선행연구자들에 의해 연구되어진 제로터의 이론적 설계 방법을 토대로 설계식의 타당성을 검증해보고 곡선족의 포락선으로 부터 로터 프로파일을 설계하는 방법을 설명하고 CAD기법을 이용한 계산과 실험을 통하여 설계 프로파일과 배제용적 산출 방법의 타당성을 비교 검증 하였다. 제로터 펌프의 특성실험을 통하여 성능을 확인하고 제로터 펌프의 디자인과 배제용적 산출 방법을 설명하였다.

• 에너지공학
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• 제17권4호
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• pp.218-226
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• 2008

가압 경수로의 주요 기기에서 발생할 수 있는 과도 2상 유동(Two-phase flow) 현상에 대한 해석을 수행하기 위해 원자로 기기 열수력 해석 코드를 개발 중에 있다. 개발중인 기기 열수력 해석 코드는 지배 방정식으로 Two-phase, three-field model을 사용하고 있으며, 복잡한 기하학적 형상의 원자로 기기를 모사하기 위해 비정렬 격자계(Unstructured grid)를 활용하고 있다. 수치해석 기법으로는, 원자로 계통 해석코드 REIAP5가 사용 중이며 대부분의 원자로 내 2상 유동 조건에서 안정적이며 정확하다고 알려진 Semi-implicit 방법을 적용하였다. 그러나 기존의 Semi-implicit 방법은 1차원, 엇갈림격자(Staggered grid)에 대해 개발되었기 때문에 이를 다차원, 비정렬, 비엇갈림 격자(Non-staggered grid)에 적용하기 위해 기존의 Semi-implicit 방법을 수정하였다. 본 논문에서는 수정된 Semi-implicit 방법을 소개하고 이를 이용해 수행한 예비 계산결과를 수록하였다.

• Ecology and Resilient Infrastructure
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• 제2권1호
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• pp.42-53
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• 2015

수제군 주변의 유동장 및 지형학적 변화에 대한 수리학적 효과를 분석하기 위하여, 본 연구는 영천강 만곡부에 설치된 저수 수제군 주변의 하도지형 특성의 변화에 대하여 모니터링과 수치모의를 수행하였다. 연구 대상지는 연장 190m로서 2008년 3월에 저수 수제군이 설치되었다. 현장 모니터링을 2008년 5월부터 2014년 4월까지 수행하였다. 설계된 수제 높이 보다 홍수시 수위가 더 높아져도 수제는 하안으로부터 흐름을 하도의 중심으로 밀어낸다. 따라서, 만곡부의 하도 안정을 위하여 설치된 수제군은 어느 정도 효과적으로 수리적 기능을 수행하고 있다. 시간의 경과에 따라 주수로가 하도의 중심으로 이동하면서 안정되었으며, 국소세굴에 의하여 수제 선단부 주변에 소가 형성되었다. 또한, 수제 사이에는 유사가 퇴적되었고, 하도에는 여울-소가 형성됨으로써 부수적으로 수제가 하도의 지형적 측면에서 다양한 물리적 특성을 창출함을 확인하였다.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.11-25
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• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.847-858
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• 2016

An overview is given on the work of the Laboratory of Nuclear Energy Systems at ETH, Zurich (ETHZ) and of the Laboratory of Thermal Hydraulics at Paul Scherrer Institute (PSI), Switzerland on tight-lattice bundles. Two-phase flow in subchannels of a tight triangular lattice was studied experimentally and by computational fluid dynamics simulations. Two adiabatic facilities were used: (1) a vertical channel modeling a pair of neighboring sub-channels; and (2) an arrangement of four subchannels with one subchannel in the center. The first geometry was equipped with two electrical film sensors placed on opposing rod surfaces forming the subchannel gap. They recorded 2D liquid film thickness distributions on a domain of $16{\times}64$ measuring points each, with a time resolution of 10 kHz. In the bubbly and slug flow regime, information on the bubble size, shape, and velocity and the residual liquid film thickness underneath the bubbles were obtained. The second channel was investigated using cold neutron tomography, which allowed the measurement of average liquid film profiles showing the effect of spacer grids with vanes. The results were reproduced by large eddy simulation + volume of fluid. In the outlook, a novel nonadiabatic subchannel experiment is introduced that can be driven to steady-state dryout. A refrigerant is heated by a heavy water circuit, which allows the application of cold neutron tomography.

• 환경영향평가
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• 제19권2호
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• pp.153-168
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• 2010

The objectives of this study are to analyze flow characteristics for a side weir, which is an inlet structure for flow discharge reduction in the main channel through 3 dimensional numerical analysis and to understand the efficiency of the overflow effect at the side weir. In this study over 40 simulations using FLOW-3D, a computational fluid dynamics program were conducted, and the results were analyzed to find the influence of the flow hydraulics, geometry, channel and weir shapes on the coefficient. It is especially considered the relatively high stage in downstream that may cause flow within channel to be backed up along the channel. Additionally by setting up the scale of simulations much larger than the existing test equipment designed by other researchers, it is intended to analyze more accurate hydraulic behavior along with the realistic hydraulic features such as structures and volumes of flow. The results show that for design with subcritical flow only if the Froude number of upstream is sustained below 0.5 and the length of weir is 33-100% of the width of channel, it is expected to improve the efficiency of the overflow over a side weir.

• 한국방재학회:학술대회논문집
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• 한국방재학회 2008년도 정기총회 및 학술발표대회
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• pp.799-802
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• 2008

General behaviors based on hydraulic characteristics of natural streams and channels have been recently analyzed and developed via various numerical models. However in the states of natural hydraulics, an experimental research must be performed simultaneously with the mathematical analysis due to effects of hydraulic properties such as meander, sediment, and so on. In this study based on 2-D advection-dispersion equation, flow and tracer experiments were performed in the S-curved meandering laboratory channel with a rectangular cross-section. The channel was equipped with instrument carriages which was equipped with an auto-traversing system to be used with velocity measuring sensors throughout the depth and breadth of the flow field. To measure concentration distribution of the salt solution was adjusted to that of the flume water by adding methanol and a red dye (KMnO4) was added to aid the visualization of the tracer cloud, the tracer was instantaneously injected into the flow as a full-depth vertical line source by the instantaneous injector and the initial concentration of the tracer was 100,000 mg/l. The secondary current as well as the primary flow pattern was analyzed to investigate the flow distribution in the meandering channels. The velocity distribution of the primary flow for all cases skewed toward the inner bank at the first bend, and was almost symmetric at the crossovers, and then shifted toward the inner bank again at the next alternating bend. Thus, one can clearly notice that the maximum velocity occurs taking the shortest course along the channel, irrespective of the flow conditions. The result of the tracer tests shows that pollutant clouds are spreading following the maximum velocity lines in each cases with various mixing patterns like superposition, separation, and stagnation of pollutant clouds. Flow characteristics in each cases performed in this study can be compared with tracer dispersion characteristics with using evaluation of longitudinal and transverse dispersion coefficients(LDC, TDC). As expected, LDC and TDC in meandering parts have been evaluated with increasing distribution and straight parts have effected to evaluate minimum of LDC and TDC due to symmetric flow patterns and attenuations of secondary flow.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1738-1748
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• 2019

Previous research has shown that the inclined condition has an impact on the natural circulation (natural circulation) mode operation of Floating Nuclear Power Plant (FNPP) mounted on the movable marine platform. Due to its compact structure, small volume, strong maneuverability, the Integral Pressurized Water Reactor (IPWR) is adopted as marine reactor in general. The OTSGs of IPWR are symmetrically arranged in the annular region between the reactor vessel and core support barrel in this paper. Therefore, many parallel natural circulation loops are built between the core and the OTSGs primary side when the main pump is stopped. and the inclined condition would lead to discrepancies of the natural circulation drive head among the OTSGs in different locations. In addition, the flow rate and temperature nonuniform distribution of the core caused by inclined condition are coupled with the thermal hydraulics parameters maldistribution caused by OTSG group operating mode on low power operation. By means of the RELAP5 codes were modified by adding module calculating the effect of inclined, heaving and rolling condition, the simulation model of IPWR in inclined condition was built. Using the models developed, the influences on natural circulation operation by inclined angle and OTSG position, the transitions between forced circulation (forced circulation) and natural circulation and the effect on natural circulation operation by different OTSG grouping situations in inclined condition were analyzed. It was observed that a larger inclined angle results the temperature of the core outlet is too high and the OTSG superheat steam is insufficient in natural circulation mode operation. In general, the inclined angle is smaller unless the hull is destroyed seriously or the platform overturn in the ocean. In consequence, the results indicated that the IPWR in the movable marine platform in natural circulation mode operation is safety. Selecting an appropriate average temperature setting value or operating the uplifted OTSG group individually is able to reduce the influence on natural circulation flow of IPWR by inclined condition.

• Geomechanics and Engineering
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• 제24권4호
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• pp.349-358
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• 2021

The Brazilian test has been widely used to determine the indirect tensile strength of rock, concrete and other brittle materials. The basic assumption for the calculation formula of Brazilian tensile strength is that the elastic moduli of rock are the same both in tension and compression. However, the fact is that the elastic moduli in tension and compression of most rocks are different. Thus, the formula of Brazilian tensile strength under the assumption of isotropy is unreasonable. In the present study, we conducted Brazilian tests on flat disk-shaped rock specimens and attached strain gauges at the center of the disc to measure the strains of rock. A tension-compression bi-modular model is proposed to interpret the data of the Brazilian test. The relations between the principal strains, principal stresses and the ratio of the compressive modulus to tensile modulus at the disc center are established. Thus, the tensile and compressive moduli as well as the correct tensile strength can be estimated simultaneously by the new formulas. It is found that the tensile and compressive moduli obtained using these formulas were in well agreement with the values obtained from the direct tension and compression tests. The formulas deduced from the Brazilian test based on the assumption of isotropy overestimated the tensile strength and tensile modulus and underestimated the compressive modulus. This work provides a new methodology to estimate tensile strength and moduli of rock simultaneously considering tension-compression bi-modularity.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2311-2320
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• 2022

The main outcomes of the experiments H2P6 performed in the thermal-hydraulics large-scale PANDA facility at PSI in the frame of the OECD/NEA HYMERES-2 project are presented in this article. The experiments of the H2P6 series consists of two PANDA tests characterized by the activation of three (H2P6_1) or one (H2P6_2) cooler(s) in an initially stratified and pressurized containment atmosphere. The initial stratification is defined by a helium-rich region located in the upper part of the vessel and a steam/air atmosphere in the lower part. The activation of the cooler(s) results i) in the condensation of the steam in the vicinity of the cooler(s), ii) the corresponding activation of large scale natural circulation currents in the vessel atmosphere, with the result of iii) the re-distribution and mixing of the Helium stratification initially located in the upper half of the vessel and iv) the continuous pressure decay. The initial helium layer represents hydrogen generated in a postulated severe accident. The main question to be answered by the experiments is whether or not the interaction of the different, localized cooler units would be important for the application of numerical methods. The paper describes the initial and boundary conditions and the experimental results of the H2P6 series with the suggestion of simple scaling laws for both experiments in terms of i) the temperature difference(s) across the cooler(s), ii) the transient steam and helium content and iii) the pressure decay in the vessel. The outcomes of this scaling indicate that the interaction between separate, closely localized units does not play a prominent role for the present experiments. It is therefore reasonable to model several units as one large component with equivalent heat transfer area and total water flow rate.