• 한국수자원학회:학술대회논문집
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• 한국수자원학회 2023년도 학술발표회
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• pp.28-28
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• 2023

Most countries worldwide are finding it difficult to make decisions regarding the utilization of water resources and the ecological flow protection of rivers because of serious water shortages and global climate warming. To overcome this difficulty, accurate ecological flow processes and protected ecological objectives are required. Since the introduction of the concept, ecological flow calculations have been developed for more than 60 years. This technical development has always been dominated by countries such as the United States, Australia, and the United Kingdom. The technical applications, however, vary substantially worldwide. Some countries, for instance, did not readjust the method because of a lack of understanding of the ecological effect or because they failed to achieve elaborate scheduling. Mostly, readjustments were not made because the users could not make their choices from among numerous methods for ecological flow. This paper presents three research results based on a systematic review of 240 methods with clear connotation boundaries. First, the ecological flow algorithm was developed along with the scientific and technological progress in the river ecosystem theory, ecohydrological relationship, and characterization and simulation of hydrological and hydrodynamic processes. In addition, the basis of the method has evolved from the hydrological process of the ecosystem, hydraulics-habitat conditions, and social development interference to whole ecosystem simulation. Second, 240 methods were classified into 50 sub-categories to evaluate their advantages and disadvantages according to the ecological flow algorithms of hydrology, hydraulics, habitat, and other comprehensive methods. According to this evaluation, 60% of the methods were not suitable for further application, including the method based on the percentage of natural runoff. Furthermore, the applicability of the remaining methods was presented according to the evaluation based on the aspects of allocation of water resources, water conservancy project scheduling, and river ecological evaluation. Third, In the future, most developing countries should strengthen the guarantee of high-standard ecological flow via a coordination mechanism for the ecological flow guarantee established under a sustainable framework or via an ecological protection pattern at the national level according to the national system. Concurrently, a reliable ecological flow demand process should also be established on the basis of detailed investigation and research on the relationship between river habitats, ecological hydrology, and ecological hydraulics. This will ensure that the real-time evaluation of ecological flow forces the water conservancy project scheduling and accurate allocation of water.

• 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.

• 한국해안해양공학회:학술대회논문집
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• 한국해안해양공학회 2002년도 한국해안해양공학발표논문집 Proceedings of Coastal and Ocean Engineering in Korea
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• pp.24-30
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• 2002

With the economic development in China, the utilization of silty and muddy coasts including the construction of deepwater harbors and channels are being carried out at a fast pace. In these projects, the key technology involved is sediment transport. Due to the complication of sediment problems under the actions of tidal currents and wind waves, physical experiments are necessary In addition to numerical model studies. (omitted)

• Nuclear Engineering and Technology
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• 제29권1호
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• pp.45-57
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• 1997

The direct contact condensation phenomenon, which occurs when steam is injected into the subcooled water, has been experimentally investigated. Two plume shapes in the stable condensation regime are found to be conical and ellipsoidal shapes depending on the steam mass flux and the liquid subcooling. Divergent plumes, however, are found when the subcooling is relatively small. The measured expansion ratio of the maximum plume diameter to the injector inner diameter ranges from 1.0 to 2.3. By means of fitting a large amount of measured data, an empirical correlation is obtained to predict the steam plume length as a function of a dimensionless steam mass flux and a driving potential for the condensation process. The average heat transfer coefficient of direct contact condensation has been found to be in the range 1.0~3.5 ㎿/$m^2$.$^{\circ}C$. Present results show that the magnitude of the average condensation heat transfer coefficient depends mainly on the steam mass fin By using dynamic pressure measurements and visual observations, six regimes of direct contact condensation have been identified on a condensation regime map, which are chugging, transition region from chugging to condensation oscillation, condensation oscillation, bubbling condensation oscillation, stable condensation, and interfacial oscillation condensation. The regime boundaries are quite clearly distinguishable except the boundaries of bubbling condensation oscillation and interfacial oscillation condensation.

• 한국전산유체공학회지
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• 제18권4호
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• pp.53-60
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• 2013

The boron transport model has been implemented into the CUPID code to simulate the boron transport phenomena of the PWR. The boron concentration conservation was confirmed through a simulation of a conceptual boron transport problem in which water with a constant inlet boron concentration injected into an inlet of the 2-dimensional vertical flow tube. The step wise boron transport problem showed that the numerical diffusion of the boron concentration can be reduced by the second order convection scheme. In order to assess the adaptability of the developed boron transport model to the realistic situation, the ROCOM test was simulated by using the CUPID implemented with the boron transportation.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.775-784
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• 2009

This paper summarizes the tests performed in the ATLAS facility during its first two years of operation (2007${\sim}$2008). Two categories of tests have been performed successfully: (a) the reflood phase of the large-break loss-of-coolant accidents in a cold leg, and (b) the breaks in one of four direct vessel injection lines. Those tests contributed to understanding the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing an evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Several important and interesting phenomena have been observed during the tests. In most cases, the ATLAS shows reasonable accident characteristics and conservative results compared with those predicted by one-dimensional safety analysis codes. A wide variety of small-break LOCA tests will be performed in 2009.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.655-676
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• 2009

The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.

• 한국전산유체공학회지
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• 제19권1호
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• pp.64-72
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• 2014

This study aimed at assessing the analysis capability of thermal-hydraulic computer code, MARS for the behaviors of the core make-up tank (CMT). The sensitivity study on the nodalization to simulate the CMT was conducted, and the MARS calculations were compared with KAIST experimental data and RELAP5/MOD3.3 calculations. The 12-node model was fixed through a nodalization study to investigate the effect of the number of nodes in the CMT (2-, 4-, 8-, 12-, 16-node). The sensitivity studies on various parameters, such as water subcooling of the CMT, steam pressure, and natural circulation flow were done. MARS calculations were reasonable in the injection time and the effects of several parameters on the CMT behaviors even though the mesh-dependency should be properly treated for reactor applications.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2013년도 춘계학술대회 논문집
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• pp.1341-1342
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• 2013

• 한국수자원학회:학술대회논문집
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• 한국수자원학회 2005년도 학술발표회(1)
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• pp.221-222
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• 2005