• Magazine of the Korean Society of Agricultural Engineers
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• v.30 no.3
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• pp.58-68
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• 1988

Most of the Korean watersheds are mountaineous and consist of various soil types and land uses And seldom watersheds are found to have long term hydrologic records. The SNUA, a hydrologic watershed model was developed to meet the unique characteristics of Korean watershed and simulate the storm hydrographs from a small mountaineous watershed. Also the applicability of the model was tested by comparing the simulated storm hydrographs and the observed from Dochuk watershed, Gwangjugun, Kyunggido The conclusions obtained in this study could be summarized as follows ; 1. The model includes the simulation of interception, evaporation and infiltration for land surface hydrologic cycle on the single storm basis and the flow routing features for both overland and channel systems. 2. Net rainfall is estimated from the continuous computation of water balance at the surface of interception storage accounting for the rainfall intensities and the evaporation losses at each time step. 3. Excess rainfall is calculated by the abstraction of infiltration loss estimated by the Green and Ainpt Model from the net rainfall. 4. A momentum equation in the form of kinematic wave representation is solved by the finite differential method to obtain the runoff rate at the exit of the watershed. 5. The developed SNUA Model is a type of distributed and event model that considers the spatial distribution of the watershed parameters and simulates the hydrograph on a single storm basis. 6. The results of verification test show that the simulated peak flows agree with the observed in the occurence time but have relative enors in the range of 5.4-40.6% in various flow rates and also show that the simulated total runoff have 6.9-32% of relative errors against the observed. 7. To improve the applicability of the model, it was thought that more studies like the application test to the other watersheds of various types or the addition of the other hydrologk components describing subsurface storages are needed.

• Journal of Ocean Engineering and Technology
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• v.24 no.1
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• pp.106-115
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• 2010

This study described the stability of riprap, which was examined by a two-dimensional physical model of a scattered riprap submarine breakwater. Artificial reef structures made of scattered riprap are used like artificial intertidal zone structures as waterfront seaside structures. To prevent topography change in such an artificial intertidal zone the energy is reduced at the scattered riprap submarine breakwater by intercepting high waves. The breaking waves are converted into flow on the front surface slope of the submarine breakwater, which follows the upper part of the artificial intertidal zone. Because of this phenomenon of resisting water flow, it is very important to calculate the required weight of the riprap to maintain its stability. The results of a physical model can be abstracted as shown below. First, distribute the wave breaking types occurring on the front surface slope of the submarine breakwater and arrange it in relation to the movement of riprap. Second, using the hydraulic phenomenon that occurs at the depth of the scattered riprap submarine breakwater, propose a calculation formula for the velocity distribution showing the influence on the stability of the riprap. Third, propose and compare values, which can be obtained by experiments and calculations for riprap stability on the front surface of the artificial intertidal zone. Fourth, calculate the required weight for riprap stability.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.426-441
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• 2003

A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

• Journal of the Society of Naval Architects of Korea
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• v.56 no.5
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• pp.410-417
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• 2019

An axisymmetric submerged body(L=5.6m, Diam=0.53m) is installed in Large Cavitation Tunnel (LCT) of KRISO and the nominal and total velocities without and with the propeller in operation, respectively, are measured using Laser Doppler Velocimeter (LDV). The flow field is nearly axisymmetric except the wake of the supporting strut, and is considered ideal to study the hydrodynamic interaction between the propeller and the oncoming axisymmetric sheared flow. The measured velocity data are then provided to compute the propeller-induced velocity to get the effective velocity, which is defined by subtracting the propeller-induced velocity from the total velocity. We adopted, in computing the induced velocity, two different methods including the vortex lattice method and the vortex tube actuator model to evaluate the resultant effective velocity distribution. To secure a fundamental base of experimental data necessary for the research on the effective wake, we measured the drag of the submerged body, the nominal and total velocity distributions at various axial locations for three different tunnel water speeds.

• The Korean Journal of Quaternary Research
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• v.22 no.1
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• pp.28-36
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• 2008

The present study has focused on the environmental changes and evidences for sedimentation in the Lake Khuvsgul catchment during the Holocene period, inferred from short core sediment (BO03) from the eastern shore of Borsog Bay, which were analyzed in order to review records of the Holocene climatic evolution and Holocene history in Northern Mongolia. For the purpose of reconstruction of natural phenomenon that occurred in the lake catchment system during the Holocene, physical and chemical properties including HCl-soluble material, biogenic silica, organic matter and grain size distribution of minerals in the core sediments have been analyzed in this study. The vertical variations in composition for these properties show distinctly that five lines of paleoenvironmental evidence occurred in the lake catchment during the Holocene. A modified age model resulting from AMS carbon dating for the BO03 core sediment shows timings of these environmental events at 9.5 Kyr BP, 8.0 Kyr BP, 5.6 Kyr BP and 3.2 Kyr BP, respectively. Paleoenvironmental changes in the Lake Khuvsgul catchment system during the Holocene highlight distinctive features of the hydrological regime and geomorphologic evolution in the lake catchment due to regional landscape and global climatic changes corresponding with the Holocene optimum and thermal optimum. In particular, the change of hydrologic regime based on the sedimentological evidence has been caused by not only overland flow due to melting water, but also base flow due to thick permafrost around Khuvsgul region.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• Journal of Advanced Marine Engineering and Technology
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• v.36 no.6
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• pp.814-822
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• 2012

To improve the flow and mixing characteristics of marine SCR system, two different mixer including up-down and swirl type mixer were considered. The purpose of this study is to analyse turbulence intensity and uniformity index in detail and to improve the performance of SCR with respect to the mixer structure. The results showed that, the concentration uniformity index is improved by about 5% with the utilization of both mixers in the front of catalyst part. Although the RMS value and relative turbulence intensity increased after the up-down type mixer, it could observed that the value of two parameters decreased with the flow proceeding forward to the downstream. For the case of swirl type mixer, the decrease of RMS value and relative turbulence intensity were relatively smaller than that of up-down type mixer, and uniform distribution of relative turbulence intensity was observed. As a results, it could be concluded that the mixing effects and the distance of the two kinds of mixer were different.

• Nuclear Engineering and Technology
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• v.51 no.7
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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.

• Journal of the Society of Naval Architects of Korea
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• v.53 no.5
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• pp.371-379
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• 2016

In the ship design process, ship motion and propulsion performance in sea waves became very important issues. Especially, prediction of ship propulsion performance during real operation is an important challenge to ship owners for economic operation in terms of fuel consumption and route-time evaluation. Therefore, it should be considered in the early design stages of the ship. It is thought that the averaged value and fluctuation of effective inflow velocity to the propeller have a great effect on the propulsion performance in waves. However, even for the nominal velocity distribution, very few results have been presented due to some technical difficulties in experiments. In this study, flow measurements near the propeller plane using a stereo PIV system were performed. Phase-averaged flow fields on the propeller plane of a KVLCC2 model ship in waves were measured in the towing tank by using the stereo PIV system and a phase synchronizer with heave motion. The experiment was carried out at fully loaded condition with making surge, heave and pitch motions free at a forward speed corresponding to Fr=0.142 (Re=2.55×106) in various head waves and calm water condition. The phase averaged nominal velocity fields obtained from the measurements are discussed with respect to effects of wave orbital velocity and ship motion. The low velocity region is affected by pressure gradient and ship motion.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.479-490
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• 2005

The thermalhydraulic performance of a CANDU-6 reactor loaded with various CANFLEX fuel bundles is evaluated by the NUCIRC code, which is incorporated with recent models of pressure drop and critical heat flux (CHF) predictions based on high-pressure steam-water tests for the CANFLEX bundle as well as a 37-element bundle. The distributions of channel flow rate, channel exit quality, critical channel power (CCP), and critical power ratio (CPR) for the CANFLEX bundles (with natural or recycled uranium fuel) in the CANDU-6 reactor fuel channel are calculated by the code. The effects of axial and radial heat flux on CCP are evaluated by assuming that the recycled uranium fuel (CANFLEX-RU) has the same geometric data as the natural uranium fuel bundle (CANFLEX-NU), but a different power distribution due to different fuel composition and refueling scheme. In addition, the effects of pressure tube creep and bearing-pad height are examined by comparing various results of uncrept, and $3.3\%\;and\;5.1\%$ crept channels loaded with CANFLEX bundles with 1.4 mm or 1.7 mm high bearing-pads with those of the 37-element bundle. The distributions of the channel flow rate and CCP for the CANFLEX-NU or -RU bundle show a typical trend for a CANDU-6 reactor channel, and the CPRs are maintained above at least 1.444 (NU) or 1.455 (RU) in the uncrept channel. The enhanced CHF of the CANFLEX bundle (particularly with 1.7mm height bearing-pads) produces a higher thermal margin and considerably less sensitivity to CCP reduction due to the pressure tube creep than the 37-element bundle. The CCP enhancement due to the raised bearing-pads is estimated to be about $3\%\~5\%$ for the CANFLEX-NU and $2\%\~6\%$ for the CANFLEX-RU bundle, respectively.