Han Young Yoon

THE CUPID CODE DEVELOPMENT AND ASSESSMENT STRATEGY

A thermal-hydraulic code, named CUPID, has been being developed for the realistic analysis of transient two-phase flows in nuclear reactor components. The CUPID code development was motivated from very practical needs, including the analyses of a downcomer boiling, a two-phase flow mixing in a pool, and a two-phase flow in a direct vessel injection system. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations are solved over unstructured grids with a semi-implicit two-step method. This paper presents an overview of the CUPID code development and assessment strategy. It also presents the code couplings with a system code, MARS, and, a three-dimensional reactor kinetics code, MASTER.

DEVELOPMENT AND PRELIMINARY ASSESSMENT OF A THREE-DIMENSIONAL THERMAL HYDRAULICS CODE, CUPID

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been developed. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations were solved over unstructured grids, which are very useful for the analysis of flows in complicated geometries. To obtain numerical solutions, the semi-implicit numerical method for the REALP5 code was modified for an application to unstructured grids, and it has been further improved for enhanced accuracy and fast running. For the verification of the CUPID code, a set of conceptual problems and experiments were simulated. This paper presents the flow model, the numerical solution method, and the results of the preliminary assessment.

MULTI-SCALE THERMAL-HYDRAULIC ANALYSIS OF PWRS USING THE CUPID CODE

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 systemscale code, MARS.

RECENT IMPROVEMENTS IN THE CUPID CODE FOR A MULTI-DIMENSIONAL TWO-PHASE FLOW ANALYSIS OF NUCLEAR REACTOR COMPONENTS

The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of a two-phase flow in light water nuclear reactor components. It can provide both a component-scale and a CFD-scale simulation by using a porous media or an open media model for a two-phase flow. In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.

Simulation of single- and two-phase natural circulation in the passive condensate cooling tank using the CUPID code

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal-hydraulic code, named CUPID, has been developed. In the present study, the CUPID code was applied for the simulation of the PASCAL test facility constructed with an aim of validating the cooling and operational performance of the passive auxiliary feedwater system (PAFS). The PAFS is one of the advanced safety features adopted in the Advanced Power Reactor + (APR+), which is intended to completely replace the conventional active auxiliary feedwater system. This paper introduces the simulation results for the passive condensate cooling tank (PCCT) of the PASCAL facility performed with the CUPID code in order to investigate the thermal-hydraulic phenomena in the PCCT. The simulation showed that the important thermal-hydraulic characteristics in the PCCT, such as two-phase natural circulation and boil-off phenomena, can be successfully reproduced by CUPID. Two important validation parameters, collapsed water level and local liquid temperature, were quantitatively well captured in the simulation. This paper presents the description of the PASCAL test facility, the physical models of the CUPID code, and its simulation result for the PCCT.

Assessment of the two-phase flow models in the CUPID code using the downcomer boiling experiment

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. We simulated the downcomer boiling experiment (DOBO) experiment in two-dimensions using the CUPID code to evaluate its two-phase flow models and verify its applicability to the downcomer boiling analysis. The simulation result showed that it can reproduce the important characteristics of the downcomer boiling, such as a flow pattern change from a bubbly flow to churn and mist flows and a circulation of liquid accelerated by bubbles. The two-phase flow models that require further improvement were identified as well for an enhanced prediction of the downcomer boiling.

Preliminary Analysis of the CANDU Moderator Thermal-Hydraulics using the CUPID Code

A transient, three-dimensional, two-phase flow analysis code, CUPID, has been developed in KAERI. In this work, we performed a preliminary analysis using the CUPID code to investigate the thermal-hydraulic behavior of the moderator in the Calandria vessel of a CANDU reactor. At first, we validated the CUPID code using the three experiments that were performed at Stern Laboratories Inc. To avoid the complexity to generate computational mesh around the Calandria tube bundles, a porous media approach was applied for the region. The pressure drop in the porous media zone was modeled by an empirical correlation. The results of the calculations showed that the CUPID code can predict the mixed flow pattern of forced and natural convection inside the Calandria vessel very well. Thereafter, the analysis was extended to a two-phase flow condition. Also, the local maximum temperature in the Calandria vessel was plotted as a function of the injection flow rate, which may be utilized to predict the local subcooling margin.

Preliminary Analysis of Two-Phase Flow in a Steam Generator Using CUPID-SG

CUPID-SG is a component scale thermal-hydraulic analysis tool based on CUPID, which is intended to be a computer code for analyzing two-phase flows in porous media with tube bundles such as PWR steam generators. CUPID-SG adopts two-fluid, three-field conservation equations, and uses an in-house semi-implicit solver to obtain numerical efficiency. CUPID-SG can handle unstructured meshes that enable its application in complex geometries. CUPID-SG also has constitutive models for a two-phase flow map, interfacial heat and mass transfer, interfacial drag, wall friction, wall heating, and heat partitioning. CUPID-SG is benchmarked against a few cases of FRIGG tests to confirm that the transport models are correctly implemented and the code is applicable to a two-phase flow over tube bundles. The comparison showed that the constitutive models are properly implemented and CUPID-SG is capable of analyzing a two-phase boiling flow over tube bundles.Copyright

Simulations of air–water flow and subcooled boiling flow using the CUPID code

This paper summarizes the physical models and a modified numerical procedure for the lift force of the CUPID code to simulate a subcooled boiling flow. A part of these physical models and the numerical procedure were verified through a calculation of air–water flow tests. A subcooled boiling flow was then calculated to assess the whole implementation. These assessments indicate that the implementation of the physical models and a modified numerical scheme are appropriate, and that the calculated primary variables of a subcooled boiling flow such as the liquid velocity and void fraction profiles can be acceptable in general, though IATE and a turbulence model for a two-phase flow are needed to be improved for a better prediction.

Assessment of the CUPID Code for Bubbly Flows in Horizontal Pipes

Abstract Two-phase flow in a horizontal pipe has a pronounced feature; that is, two-phase-flow parameters are highly nonsymmetric because gravity is perpendicular to the mean flow direction. Thus, three-dimensional analysis is necessary for the accurate prediction of two-phase flow in a horizontal pipe, such as the hot leg and cold leg of a pressurized water reactor and the pressure tubes in a CANDU reactor. In this study, we simulated bubbly flows in horizontal pipes using the CUPID code, which adopts a two-fluid, three-field model for two-phase flow. In the preliminary calculations, it was found that the particle-averaged two-fluid momentum equation, rather than the standard two-fluid momentum equation, predicts a physically reasonable slip ratio and nondrag forces, except turbulent dispersion forces have negligible effects on the radial void distribution when the particle-averaged two-fluid momentum equation is used. Based on the results, we selected the physical models and computational mesh for subsequent code assessment using various bubbly flow experiments in horizontal pipes. The turbulent dispersion force model was improved to take into account the large void fraction change at the top. The results of the code assessment show good predictions for the axial pressure drop, liquid velocity, and turbulent kinetic energy profile and predict reasonably well the effects of jl and jg on two-phase-flow parameters. However, additional studies are needed for more accurate prediction of the nonsymmetric distribution of gas velocity and turbulent kinetic energy.