• International Journal of Fluid Machinery and Systems
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• 제8권4호
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• pp.318-326
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• 2015

A valve-less electro-hydraulic servo steering device (short: VSSD) for ships was chosen as a study object, and its mathematic model of hydraulic shock was established on the basis of flow properties and force balance of each component. The influence of system structure parameters, changing rate of motor speed and external load on hydraulic shock strength was simulated by the method of numerical simulation. Experiment was designed to test the hydraulic shock mathematic model of VSSD. Experiment results verified the correctness of the model, and the model provided a correct theoretical method for the calculation and control of hydraulic shock of valve-less electro-hydraulic servo steering device.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2004년도 유체기계 연구개발 발표회 논문집
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• pp.473-478
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• 2004

This work presents the Thermal Hydraulic analysis has been performed for a 19-pin wire-spacer fuel assembly using three-dimensional Reynolds-averaged Navier-Stokes equations. SST model is used as a turbulence closure. The whole fuel assembly has been analyzed for one period of the wire-spacer using periodic boundary condition at inlet and outlet of the calculation domain. The overall results far a preliminary calculation show a good agreement with the experimental observations. It has been found that the major unidirectional flows are the axial velocity in sub-channels and the peripheral sweeping flows and the velocities are found to be following a cyclic path of period equal to the wire-wrap pitch. The temperature is found to be maximum in the central region and also, there exist a radial temperature gradient between the fuel rods. The major advantage of performing this kind of analysis is the prediction of thermal-hydraulic behavior of a fuel assembly with much ease.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.831-846
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• 2012

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1721-1728
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• 2019

The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback, reconstruct the pin-by-pin power. It has lots of functions such as changing core status calculation, critical searching, control rod value calculation, coefficient calculation and so on. The main theory and functions of CORCA-3D code are introduced and validated with a lot of reactor measured data and the SCIENCE system. Now, CORCA-3D code has been applied in ACP type reactor nuclear cores design.

• 한국농공학회지
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• 제11권4호
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• pp.1832-1840
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• 1969

• Geomechanics and Engineering
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• 제7권4호
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• pp.403-430
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• 2014

Shale gas formations exhibit strong mechanical and strength anisotropies. Thus, it is necessary to study the effect of anisotropy on the hydraulic fracture initiation pressure. The calculation model for the in-situ stress of the bedding formation is improved according to the effective stress theory. An analytical model of the stresses around wellbore in shale gas reservoirs, in consideration of stratum dip direction, dip angle, and in-situ stress azimuth, has been built. Besides, this work established a calculation model for the stress around the perforation holes. In combination with the tensile failure criterion, a prediction model for the hydraulic fracture initiation pressure in the shale gas reservoirs is put forward. The error between the prediction result and the measured value for the shale gas reservoir in the southern Sichuan Province is only 3.5%. Specifically, effects of factors including elasticity modulus, Poisson's ratio, in-situ stress ratio, tensile strength, perforation angle (the angle between perforation direction and the maximum principal stress) of anisotropic formations on hydraulic fracture initiation pressure have been investigated. The perforation angle has the largest effect on the fracture initiation pressure, followed by the in-situ stress ratio, ratio of tensile strength to pore pressure, and the anisotropy ratio of elasticity moduli as the last. The effect of the anisotropy ratio of the Poisson's ratio on the fracture initiation pressure can be ignored. This study provides a reference for the hydraulic fracturing design in shale gas wells.

• 한국도로학회논문집
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• 제15권3호
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• pp.45-52
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• 2013

PURPOSES : This study is to construct the regression models of drainage asphalt concrete specimens and to provide the appropriate coefficients of hydraulic conductivity prediction models. METHODS: In terms of easy calculation of the hydraulic conductivity from porosity of asphalt concrete pavement, the estimation model of hydraulic conductivity was proposed using regression analysis. 10 specimens of drainage asphalt concrete pavement were made for measurement of the hydraulic conductivity. Hydraulic conductivity model proposed in this study was calculated by empirical model based on porosity and the grain size. In this study, it shows the compared results from permeability measured test and empirical equation, and the suitability of proposed model, using regression analysis. RESULTS: As the result of the regression analysis, the hydraulic conductivity calculated from the proposal model was similar to that resulted from permeability measured test. Also result of RMSE (Root Mean Square Error) analysis, a proposed regression model is resulted in more accurate model. CONCLUSIONS: The proposed model can be used in case of estimating the hydraulic conductivity at drainage asphalt concrete pavements in fields.

• Nuclear Engineering and Technology
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• 제40권3호
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• pp.199-212
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• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

• 한국안전학회지
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• 제36권2호
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.357-362
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• 1996

LOFT LBLOCA test, LP-02-6 was analyzed using RELAP5/MOD3.2. It has a distinguished thermal-hydraulic phenomenon of a positive bottom-up core flow in tile blowdown phase. A modified nodalization which is based on that used in LP-LB-1 calculation by Lubbesmeyer was used in the calculation. RELAP5/MOD3.2 predicted overall system hydraulic behavior relatively well. However, the bottom-up quenching in the middle part of the core was not predicted sufficiently. It was demonstrated also that the peak cladding temperature can be predicted well by adjusting a discharge coefficient. But more improvements are needed in order to apply this code to actual plants with less user dependency.