Hyungrae Kim

EXPERIMENTAL INVESTIGATIONS ON HEAT TRANSFER TO CO₂ FLOWING UPWARD IN A NARROW ANNULUS AT SUPERCRITICAL PRESSURES

Heat transfer experiments in an annulus passage were performed using SPHINX(Supercritical Pressure Heat Transfer Investigation for NeXt Generation), which was constructed at KAERI(Korea Atomic Energy Research Institute), to investigate the heat transfer behaviors of supercritical . was selected as the working fluid to utilize its low critical pressure and temperature when compared with water. The mass flux was in the range of 400 to 1200 and the heat flux was chosen at rates up to 150 . The selected pressures were 7.75 and 8.12 MPa. At lower mass fluxes, heat transfer deterioration occurs if the heat flux increases beyond a certain value. Comparison with the tube test results showed that the degree of heat transfer deterioration in the heat flux was smaller than that in the tube. In addition, the Nusselt number correlation for a normal heat transfer mode is presented.

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.

Experimental Investigation on the Heat Transfer Characteristics in Upward Flow of Supercritical Carbon Dioxide

Abstract The SuperCritical Water-cooled Reactor (SCWR) is one of the candidates for the fourth-generation nuclear power plant, and it uses light water as a coolant. Heat transfer between a fuel assembly and water may degrade at certain conditions of supercritical pressure flows. Therefore, accurate and reliable estimation of heat transfer coefficients is necessary for the design of the fuel assembly and the reactor core. A series of heat transfer tests has been carried out at a test facility named SPHINX by using carbon dioxide as a stimulant of water. The tests produced heat transfer data of the supercritical pressure flows inside a circular tube of 4.4-mm inside diameter at varying operating pressures, mass fluxes, and wall heat fluxes. The test range of the mass flux was 400 to 1200 kg/m2 s, and the maximum heat flux was 150 kW/m2. The operating pressures were 7.75, 8.12, and 8.85 MPa in the tests. The test results were compared with estimations of the existing correlations for supercritical pressure flows. The comparison showed reasonable agreement between our data and the correlations. However, a rather large departure from the normal heat transfer correlations was observed near pseudocritical temperatures. Besides the comparison of the normal heat transfer coefficients, the onset of heat transfer deterioration was compared between the test cases and two existing criteria. One of the criteria was derived from experiments by using Freon but with a test section of identical geometry while the other criterion was for a flow of carbon dioxide in a larger bore channel than ours. Both criteria showed fair agreement with our experiment. Most test cases with noticeable heat transfer degradation were located in the region of deterioration predicted by the criteria.

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

Porous Media Approach of a CFD Code to Analyze Thermal Hydraulics of PWR Components

This paper presents a set of numerical procedure to innovate CFD code into a PWR component analysis code. A porous media approach is adapted to two-fluid model and conductor model, and a pack of constitutive relations close the numerical model into a PWR component analysis code. The separate verification calculations on conductor model and porous media approach, and the validation calculation for the integrated component-scale code are introduced.Copyright