• Journal of computational fluids engineering
• /
• v.15 no.2
• /
• pp.86-94
• /
• 2010

A thermal-hydraulic code, named CUPID, has been developed for the analysis of transient two-phase flows in nuclear reactor components. A two-fluid three-field model was used for steam-water two-phase flows. To obtain numerical solutions, the finite volume method was applied over unstructured cell-centered meshes. In steam-water two-phase flows, a phase change, i.e., evaporation or condensation, results in a great change in the flow field because of substantial density difference between liquid and vapor phases. Thus, two-phase flows are very sensitive to the local pressure distribution that determines the phase change. This in turn puts emphasis on the accurate evaluation of local pressure gradient. This paper presents a new reconstruction method to evaluate the pressure gradient at cell centers on unstructured meshes. The results of the new scheme for a simple test function, a gravity-driven cavity, and a wall boiling two-phase flow are compared with those of the previous schemes in the CUPID code.

• Transactions of the Korean Society of Mechanical Engineers
• /
• v.10 no.3
• /
• pp.326-334
• /
• 1986

The effects of flow interaction between adjacent sprays in twin-spray system on the spatial distribution of injected liquid (water) and drop size distribution in condensable (steam) environment were carefully observed through experiments. The spatial distribution of injected liquid in twin-spray system appears to be more uniform than the simple superposition of the spatial distributions of liquid obtained from each individual spray. Drop size distribution was obtained by using the immersion sampling technique. It was found that, in the twin-spray, the larger numbers of small drops are collected throughout the spraying region due to the increase of entrainment velocity of ambient steam compared with the case of simple superposition of each individual spray. Moreover, in the overlapped portion of the twin-spray, the drop size distribution was changed also due to the collision between large drops. As a result, the behavior of twin-spray system (and eventually multiple-spray system) can not be predicted precisely by simple superposition of the behaviors of each constituting spray. Hence, for the design of multiple spray system, the effect of flow interaction between sprays should be taken into account seriously.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.3 no.5
• /
• pp.337-349
• /
• 1991

Heat and mass transfer rates to spray water droplets for spray transients in a high pressure vessel have been predicted by two different droplet models: the complete mixing model and the non-mixing model. In this process, the ambient fluid surrounding the droplets is a real-gas mixture composed of saturated steam and noncondensable hydrogen gas at high pressure. The physical properties of the mixture are estimated by applying the concept of compressibility factor and using appropriate correlations. A computer program, DROPHMT, to calculate the heat and mass transfer rates for two different droplet models has been developed. As an illustrative application of the computer program to engineering practices, heat and mass transfer rates to spray water droplets for spray transients in a Pressurized Water Reactor (PWR) pressurizer have been calculated, and the typical results have been provided.

• 한국전산유체공학회:학술대회논문집
• /
• 2010.05a
• /
• pp.380-388
• /
• 2010

A thermal-hydraulic code, named CUPID, has been developed for the analysis of transient two-phase flows in nuclear reactor components. A two-fluid three-field model was used for steam-water two-phase flows. To obtain numerical solutions, the finite volume method was applied over unstructured cell-centered meshes. In steam-water two-phase flows, a phase change, i.e., evaporation of condensation, results in a great change in the flow field because of substantial density difference between liquid and vapor phases. Thus, two-phase flows are very sensitive to the local pressure that determines the phase change. This in turn puts emphasis on the accurate evaluation of local pressure gradient. This paper presents a new numerical scheme to evaluate the pressure gradient at cell centers on unstructured meshes. The results of the new scheme for a simple test function a gravity-driven cavity, and a wall boiling two-phase flow are compared with those of the previous schemes in the cupid code.

• Nuclear Engineering and Technology
• /
• v.34 no.5
• /
• pp.421-432
• /
• 2002

As one of the advanced design features of the APR1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, test results of direct ECC bypass performed in the steam-water test facility tailed MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) are presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’with the l/4.93 length scale . From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region are obtained.

• Journal of the Korean Society of Fisheries and Ocean Technology
• /
• v.34 no.2
• /
• pp.212-222
• /
• 1998

Heat transfer performance improvement by fin and groovs is studied for condensation of R-11 on integral-fin tubes. Eight tubes with trapczodially shaped integral-fins having fin density from 748 to 1654fpm(fin per meter) and 10, 30 grooves are tested. A plain tube having the same diameter as the finned tubes is also used for comparison. R-11 condensates at saturation state of 32 $^{\circ}C$ on the outside tube surface coded by inside water flow. All of test data are taken at steady state. The heat transfer loop is used for testing singe long tubes and cooling is pumped from a storage tank through filters and folwmeters to the horizontal test section where it is heated by steam condensing on the outside of the tubes. The pressure drop across the test section is measured by menas pressure gauge and manometer. The results obtained in this study is as follows : 1. Based on inside diameter and nominal inside area, overall heat transfer coefficients of finned tube are enhanced up to 1.6 ~ 3.7 times that of a plain tube at a constant Reynolds number. 2. Friction factors are up to 1.6 ~ 2.1 times those of plain tubes. 3. The constant pumping power ratio for the low integral-fin tubes increase directly with the effective area to the nominal area ratio, and with the effective area diameter ratio. 4. A tube having a fin density of 1299fpm and 30 grooves has the best heat transfer performance.

• Nuclear Engineering and Technology
• /
• v.52 no.1
• /
• pp.59-68
• /
• 2020

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.27 no.8
• /
• pp.1051-1060
• /
• 2003

The heat and mass transfer on two kinds of tube surfaces (bare stainless steel tube and Teflon coated tube) in steam-air mixture flow are experimentally studied to obtain design data for the heat exchanger of the latent heat recovery from flue gas. In the test section, 3-tubes are horizontally installed, and steam-air mixture is vertically flowed from the top to the bottom. The pitch between tubes is 67mm, the out-diameter of tube is 25.4mm, and the thickness is 1.2mm ; blockage factor (cross sectional tube area over the cross sectional area of the test section) is about 0.38. All of sensors and measurement systems (RTD, pressure sensor, flow-meter, relative humidity sensor, etc.) are calibrated with certificated standard sensors and the uncertainty for the heat transfer measurement is surveyed to have the uncertainty within 7%. As experimental results, overall heat transfer coefficient of the Teflon (FEP) coated tube is degraded about 20% compared to bare stainless tube. The degradation of overall heat transfer coefficient of Teflon coated tube comes from the additional heat transfer resistance due to Teflon coating. Its magnitude of heat transfer resistance is comparable to the in-tube heat transfer resistance. Nusselt and Sherwood numbers on Teflon (FEP) coated surface and bare stainless steel surface are discussed in detail with the contact angles of the condensate.

• Nuclear Engineering and Technology
• /
• v.36 no.1
• /
• pp.24-35
• /
• 2004

In order to analyze the hydrogen distribution during a severe accident in the APR1400 containment, GASFLOW II was used. For the APR1400 NPP, a hydrogen mitigation system is considered from the design stage, but a fully time-dependent, three-dimensional analysis has not been performed yet. In this study GASFLOW code II is used for the three-dimensional analysis. The first step to analysis involving hydrogen behavior in a full containment with the GASLOW code is to generate a realistic geometry model, which includes nodalization and modeling of the internal structures such as walls, ceilings and equipment. Geometry modeling of the APR1400 is conducted using GUI program by overlapping the containment cut drawings in a graphical file format on the mesh view. The total number of mesh cells generated is 49,476. And the calculated free volume of the APR1400 containment by GASFLOW is almost the same as the value from the GOTHIC modeling. A hypothetical SB-LOCA scenario beyond design base accident was selected to analyze the hydrogen behavior with the hydrogen mitigation system. The source of hydrogen and steam for the GASFLOW II analysis is obtained from a MAAP calculation. Combustion pressure and temperature load possibilities within the compartments used in the GOTHIC analysis are studied based on the Sigma-Lambda criteria. Finally the effectiveness of HMS installed in the APR1400 containment is evaluated from the point of severe accident management

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2311-2320
• /
• 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.