• Journal of computational fluids engineering
• /
• v.19 no.2
• /
• pp.24-29
• /
• 2014

In this study, a parallel performance of CUPID-code has been investigated for both Ethernet and Infiniband network system to examine the effect of cache memory and network-speed. Bi-conjugate gradient solver of CUPID-code has been parallelised by using domain decomposition method and message passing interface (MPI). It is shown that the parallel performance of Ethernet-network system is worse than that of Infiniband-network system due to the slow network-speed and a small cache memory. It is also found that the parallel performance of each system deteriorates for a small problem due to the communication overhead, but the performance of Infiniband-network system is better than Ethernet-network system due to a much faster network-speed. For a large problem, the parallel performance depends less on network system.

• Journal of computational fluids engineering
• /
• v.18 no.4
• /
• pp.53-60
• /
• 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.

• Journal of Energy Engineering
• /
• v.23 no.4
• /
• pp.95-105
• /
• 2014

The CUPID code has been developed for a transient, three-dimensional, two-phase flow analysis at a component scale. It has been validated against a wide range of two-phase flow experiments. Especially, to assess its applicability to single- and two-phase flow analyses in the Calandria vessel of a CANDU nuclear reactor, it was validated using the experimental data of the 1/4-scaled facility of a Calandria vessel at the STERN laboratory. In this study, a preliminary thermal-hydraulic analysis of the CANDU reactor moderator tank using the CUPID code is carried out, which is based on the results of the previous studies. The complicated internal structure of the Calandria vessel and the inlet nozzle was modeled in a simplified manner by using a porous media approach. One of the most important factors in the analysis was found to be the modeling of the tank inlet nozzle. A calculation with a simple inlet nozzle modeling resulted in thermal stratification by buoyance, leading to a boiling from the top of the Calandria tank. This is not realistic at all and may occur due to the lack of inlet flow momentum. To improve this, a new nozzle modeling was used, which can preserve both mass flow and momentum flow at the inlet nozzle. This resulted in a realistic temperature distribution in the tank. In conclusion, it was shown that the CUPID code is applicable to thermal-hydraulic analysis of the CANDU reactor moderator tank using the cost-effective porous media approach and that the inlet nozzle modeling is very important for the flow analysis in the tank.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.40 no.7
• /
• pp.457-470
• /
• 2016

Most CFD codes, that mainly adopt the heat partitioning model as the wall boiling model, have shown low accuracies in predicting the two-phase flow parameters of subcooled boiling phenomena under low pressure conditions. In this study, a number of subcooled boiling experiments in vertical channels were analyzed using a thermal-hydraulic component code, CUPID. The prediction of the void fraction distribution using the CUPID code agreed well with experimental data at high-pressure conditions; whereas at low-pressure conditions, the predicted void fraction deviated considerably from measured ones. Sensitivity tests were performed on the submodels for major parameters in the heat partitioning model to find the optimized sets of empirical correlations suitable for low-pressure subcooled flow boiling. The effect of the K-factor on the void fraction distribution was also evaluated.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.54-67
• /
• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Nuclear Engineering and Technology
• /
• v.47 no.5
• /
• pp.567-578
• /
• 2015

In a nuclear reactor containment, wall condensation forms with noncondensable gases and their accumulation near the condensate film leads to a significant reduction in heat transfer. In the framework of nuclear reactor safety, the film condensation in the presence of noncondensable gases is of high relevance with regards to safety concerns as it is closely associated with peak pressure predictions for containment integrity and the performance of components installed for containment cooling in accident conditions. In the present study, CUPID code, which has been developed by KAERI for the analysis of transient two-phase flows in nuclear reactor components, is improved for simulating film condensation in the presence of noncondensable gases. In order to evaluate the condensate heat transfer accurately in a large system using the two-fluid model, a mass diffusion model, a liquid film model, and a wall film condensation model were implemented into CUPID. For the condensation simulation, a wall function approach with a heat/mass transfer analogy was applied in order to save computational time without considerable refinement for the boundary layer. This paper presents the implemented wall film condensation model, and then introduces the simulation result using the improved CUPID for a conceptual condensation problem in a large system.

• Journal of computational fluids engineering
• /
• v.14 no.3
• /
• pp.86-94
• /
• 2009

The non-conservative form of momentum equations is often used for some two-phase flow codes instead of a conservative form because of numerical convenience. Another non-conservative form, so called, a semi-conservative form can improve the numerical solution of these codes maintaining the numerical convenience. It is close to the conservative form but still maintains the feature of the non-conservative form. A semi-conservative form of the momentum equations and a non-conservative form of the momentum equations are implemented in CUPID[1] code. The numerical results of the semi-conservative and the non-conservative forms are compared against analytical solutions and the solutions of the FLUENT code that uses the conservative form. The results clearly showed that the semi-conservative form of the momentum equations provides better solutions than the non-conservative form, especially for heterogeneous two-phase flows.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.38 no.1
• /
• pp.71-79
• /
• 2014

A parallelization of the bi-conjugate gradient solver for the pressure equation of the CUPID (component unstructured program for interfacial dynamics) code, which was developed for analyzing the components of a pressurized water-cooled reactor, was studied in a symmetric multi-processing system. The parallel performance was investigated for three typical parallel programming models (MPI, OpenMP, Hybrid) by solving incompressible backward-facing step flow at various grid resolutions. It was confirmed that parallel performance was low when problem size was small or the memory requirement for each thread was considerably higher than the cache memory. Furthermore, it was shown that MPI was better than OpenMP regardless of the problem size, and Hybrid was the best when the number of threads was relatively small.

• Journal of computational fluids engineering
• /
• v.14 no.4
• /
• pp.13-22
• /
• 2009

A two-phase (gas and liquid) flow analysis solver, named CUPID, has been developed for a realistic simulation of transient two-phase flows in light water nuclear reactor components. In the CUPID solver, a two-fluid three-field model is adopted and the governing equations are solved on unstructured grids for flow analyses in complicated geometries. For the numerical solution scheme, the semi-implicit method of the RELAP5 code, which has been proved to be very stable and accurate for most practical applications of nuclear thermal hydraulics, was used with some modifications for an application to unstructured non-staggered grids. This paper is concerned with the effects of interpolation schemes on the simulation of two-phase flows. In order to stabilize a numerical solution and assure a high numerical accuracy, the second-order upwind scheme is implemented into the CUPID code in the present paper. Some numerical tests have been performed with the implemented scheme and the comparison results between the second-order and first-order upwind schemes are introduced in the present paper. The comparison results among the two interpolation schemes and either the exact solutions or the mesh convergence studies showed the reduced numerical diffusion with the second-order scheme.

• Nuclear Engineering and Technology
• /
• v.53 no.8
• /
• pp.2488-2498
• /
• 2021

The wall film-wise condensation plays an important role in the heat transfer processes of heat exchangers, refrigerators, and air conditioner. In the field of nuclear engineering, steam condensation is often utilized in safety systems to remove the core decay heat under both transient and accident conditions. In particular, passive containment cooling system (PCCS), are designed to ensure containment safety under severe accident conditions. A computational fluid dynamics (CFD) scale analysis has been conducted to calculate the heat transfer rate of the PCCS. However, despite the increase in computing power, there are challenges in the long-term transient simulation of containment using CFD scale codes. In this study, a heat structure coupling between the CFD and system analysis codes was performed to efficiently analyze PCCS. In addition, the component unstructured program for interfacial dynamics (CUPID) was improved to analyze the condensation behavior of ternary gas mixtures. Thereafter, the condensation heat transfer on the primary side was calculated using the improved CUPID and CFD code, whereas that on the secondary side was simulated using MARS. Both the coupled codes were validated against the CONAN facility database. Finally, conjugate heat transfer simulations with wall condensation in the presence of non-condensable gases were appropriately performed.