Jin Ho Song

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.

Heat Transfer Test in a Vertical Tube Using CO2 at Supercritical Pressures

Heat transfer test facility, SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), was constructed at KAERI (Korea Atomic Energy Research Institute) for an investigation of the thermal-hydraulic behaviors of supercritical CO2 at the various geometries of the test section. The test data will be used for the reactor core design of the SCWR (SuperCritical Water-cooled Reactor). As a working fluid, CO2 was selected to make use of the low critical pressure and temperature of CO2 compared with water. An experimental study was carried out in the SPHINX to investigate the characteristics of heat transfer and pressure drop at a vertical single tube with an inside diameter of 4.4mm in case of an upward flow of supercritical CO2. The heat and mass fluxes were varied at a given pressure. The mass flux was in the range of 400~1;200 kg/m2 s and the heat flux was chosen up to 150 kW/m2. The selected pressures were 7.75, 8.12, and 8.85 MPa. A heat transfer deterioration occurred at the lower mass fluxes. The experimental heat transfer coefficients were compared with the ones predicted by several existing correlations. The standard deviation was about 20% for each correlation and an apparent discrepancy was not found among the correlations. The major components of the pressure drop were a gravitational pressure drop and a frictional pressure drop. The frictional pressure drop increases as the mass flux and heat flux increase.

Insights from the Recent Steam Explosion Experiments in TROI

The paper discusses the results of steam explosion experiments of TROI-13, TROI-14, and TROI-15, which were performed under the research program named “Test for Real cOrium Interaction with water (TROI).” TROI-13 and TROI-14 used corium, which is a mixture of UO2 and ZrO2 at a 70:30wt%, while TROI-15 used ZrO2. These three cases resulted in spontaneous steam explosions. It is an important observation from the aspect of the explosivity of prototypic material, as it was not observed in the previous experiments. Various aspects of the test results including cold crucible melting, hydrogen generation, melt temperature measurement, debris morphology, size distribution of the debris, dynamic pressure, dynamic force, shape of the melt jet, location of the trigger, response of the vessel pressure, and the response of the water pool temperature are discussed. The potential contributors to the explosivity of corium are suggested.

INVESTIGATIONS ON THE RESOLUTION OF SEVERE ACCIDENT ISSUES FOR KOREAN NUCLEAR POWER PLANTS

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

OVERVIEW OF CONTAINMENT FILTERED VENT UNDER SEVERE ACCIDENT CONDITIONS AT WOLSONG NPP UNIT 1

Containment Filtered Vent Systems (CFVSs) have been mainly equipped in nuclear power plants in Europe and Canada for the controlled depressurization of the containment atmosphere under severe accident conditions. This is to keep the containment integrity against overpressure during the course of a severe accident, in which the radioactive gas-steam mixture from the containment is discharged into a system designed to remove the radionuclides. In Korea, a CFVS was first introduced in the Wolsong unit-1 nuclear power plant as a mitigation measure to deal with the threat of over pressurization, following post-Fukushima action items. In this paper, the overall features of a CFVS installation such as risk assessments, an evaluation of the performance requirements, and a determination of the optimal operating strategies are analyzed for the Wolsong unit 1 nuclear power plant using a severe accident analysis computer code, ISAAC.

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.

The Effect of Material Composition on the Strength of a Steam Explosion

To investigate the fundamental mechanism behind the recent experimental observation that the composition of a material considerably affects the strength of a steam explosion, physical and chemical analyses for the fast-quenched particles of a prototypic corium were performed. Six cases including fully oxidized and partially oxidized corium were selected for the study, in which the melt composition was changed, while the other initial and boundary conditions of the molten fuel and water interaction tests, such as the melt temperature, amount of water, and free fall height, were maintained the same. The proposition that the inner structure and solidification behavior of particles due to the existence of a mushy phase are responsible for the strong influence of the material composition on the strength of a steam explosion was examined from the results of physical and chemical analysis. The results of the present analysis are supportive of the proposed argument.

A Scaling Analysis for a Filtered Containment Venting System

Abstract A scaling method is proposed for the design of a reduced-scale experimental facility for testing the performance of a newly proposed filtered containment venting system (FCVS). A full-height facility at prototypic pressure and temperature conditions is chosen to preserve the fundamental physics such as depressurization rate, two-phase mixture level, and scrubbing process. The geometrical similarities in terms of the ratio of the cross-sectional area and geometric and frictional loss coefficient are preserved for each component in the FCVS. Scaling of the number of components in the reduced-scale test facility is suggested using the prototypic components of the FCVS including a venturi scrubber, a cyclone, a metal fiber filter, and a molecular sieve. This approach minimizes scaling distortions. A properly scaled test facility allows testing in a wide range of initial and boundary conditions such that it can predict the full performance of the prototypic FCVS.

A Core Catcher Concept and First Experimental Results

In a postulated core melt accident, if a molten core is released outside a reactor vessel despite taking mitigation actions, the core debris would relocate in the reactor cavity region and attack the concrete wall and basemat of the reactor cavity. This will potentially result in inevitable concrete decompositions and possible radiological releases. To prevent direct contact of the melt and basemat concrete of the cavity, a core catcher concept is suggested, which can passively arrest and stabilize the molten core material inside the reactor cavity. The core catcher system includes a retention device for the molten core material, a cooling water storage tank, and a compressed gas tank. Upon ablation of the sacrificial layer on top of the retention device while molten core material is discharged, a mixture of water and gas is injected from below. It is expected that a simultaneous injection of water and gas could prevent a possible steam explosion/spike. It could also suppress the rapid release of steam which might result in fast over-pressurization of the containment. A test facility for the core catcher using a thermite reaction technique for the generation of the melt was designed and constructed at KAERI. The first series of tests were performed by using a mixture of Al, Fe2 O3 , and CaO as a stimulant. As a first try, only water was injected from the bottom of the melt through five water injection nozzles when the melt front reached the water injection nozzles. In this paper, the core catcher concept and the related provisions are suggested. A description of the test facility for the core catcher, the thermite composition, and the methods of experiment is included. The first experimental results with only water injected from the bottom of the melt are discussed.Copyright

THERMAL HYDRAULIC PHENOMENA IN A WATER POOL DUE TO AIR BUBBLE OSCILLATION

In both a boiling water reactor and an advanced pressurized water reactor such as the one currently under construction in Korea named APR1400, when a pressure relieving system is in operation, water, air and steam discharge successively into a sub-cooled water pool through spargers. Among the phenomena occurring during the discharging processes, the air bubble clouds with a low-frequency and high-amplitude oscillation may result in significant damage to the submerged structures if a resonance between the bubble clouds and structures occur. This study deals with a numerical prediction of the pressure field generated by the oscillation of air bubbles. The analysis was performed by using a commercial thermal hydraulic analysis code, FLUENT, version 4.5. The multiphase flows of water, air and steam were simulated by the VOF (Volume Of Fluid) model contained in the code. Unlike Kim et al. [1] , the LRR (Load Reduction Ring) of the sparger is artificially blocked for the investigation of LRR effects on the pressure field. It also includes the effects of air mass and inlet pressure in the piping on the pressure field.