[en] This report contains two analysis models on the condensation of stable steam jets discharging into a quenching tank with subcooled water from a single horizontal pipe. The analysis results are compared with the conventional experimental ones which were performed to measure the shapes and lengths of condensing steam jet, the temperatures of the flow field and the condensation heat transfer coefficients. The first analysis model is a numerical one to analyse the condensing steam jet by employing the locally homogeneous flow approximation of two phase flow in conjunction with a k- epsilon -g ∼ model of turbulence properties. In this model, the turbulence is represented by differential equations for its kinetic energy and dissipation. A differential equation for the concentration fluctuations is solved and a clipped normal probability distribution function is proposed for the mixture fraction. The second analysis model is an analytic one to predict the configuration of condensing steam jet cavity by employing the mass, momentum and energy equations as well as a thermal balance equation with condensing characteristics at the steam/water interface for the axi-symmetric coordinates. In this model, the extremely large heat transfer rate at the steam/water interface is reflected in the effective thermal conductivity estimated from the conventional experimental results. The proposed analysis models are evaluated comparing the analysis results with the experimental ones. It shows that both results are in good agreement. In summary, this report can be used for the development of the analysis model on the complex sparger, since it proposes the analysis models on the condensing steam jet discharging from a single pipe which is considered a simple sparger

[en] If the safety depressurization and vent system of APR-1400, the Korean next generation reactor, is in operation, water, air and steam are successively discharging into an in-containment refueling water storage tank through a sparger. Among the phenomena occurring during the discharging processes, the air bubble clouds produce a low-frequency and high-amplitude oscillatory loading, which may result in the most significant damages to the submerged structures if the oscillation frequency is the same or close to the natural frequency of the structures. The involved phenomena are so complicated that most of the prediction of frequency and pressure loads has been resorted to experimental work and computational approach has been precluded. Thus, it is valuable to develop a computational model on the air bubble cloud oscillation, whose loads should be considered in the design of sparger and submerged structures. This report deals with a numerical simulation on the behavior of air bubble clouds discharging into a water pool through a sparger, by using a commercial thermal hydraulic analysis code, FLUENT, version 4.5. Among the multiphase flow models, the VOF(Volume Of Fluid) model was selected to simulate the water, air and steam flows. A satisfactory result was obtained comparing the analysis results with the ABB-atom test results which had been performed for the development of sparger. In addition, effects of air mass and inlet condition of the pipe on the behavior of air bubble cloud were included. It was found that the oscillation phases of two air bubble clouds formed at the LRR and sparger head have an impact on the pressure field in the pool

[en] The Korea Next Generation Reactor(APR-1400) provides an IRWST(In-containment Refueling Water Storage Tank) and SDVS(Safety Depressurization and Vent System) including POSRVs(Pilot Operated Safety Relief Valve) and spargers in order to enhance its safety. In case an accident of opening the POSRV occurs, the steam with high pressure and temperature is discharged through the spargers attached at the end of the pipings of the SDVS. Before the steam is discharged, water and air existing inside the piping are discharged consequently. Experimentally and analytically, it is well known that the discharged air with low frequency oscillation produces dynamic loads which cause severe impacts on the IRWST structures. This report contains a pressure forcing function of bubble cloud and a bubble radius time history for APR-1400, based on the wall dynamic pressures measured in the unit cell test with a APR-1400 sparger. The thermal hydraulic parameters affecting the maximum bubble cloud pressure were compared between the unit cell test conditions and the APR-1400 design data and the maximum bubble cloud pressure of APR-1400 was calculated. The wall effects of the unit cell test on the pressure forcing function were analyzed using a Rayleigh-Plesset equation with a mirror image method of multiple bubbles. The design data in this report are used for a load analysis of IRWST wall and submerged structures by the AE

[en] The size of a sub-channel of the conceptual SCWR core design studied at KAERI is 6.5 mm. In order to provide heat transfer information in such a narrow sub-channel at supercritical pressure, an experiment was performed with a test section made of Inconel 625 tube of 6.32 mm ID. The test pressures were 7.75 and 8.12 MPa corresponding to 1.05 and 1.1 times the critical pressure of CO₂, respectively. The mass flux and heat flux, which were in the range of 285 ∼ 1200 kg/m2s and 30 ∼ 170 kW/m2, were changed at a given system pressure. The corresponding Reynolds numbers are 1.8 x 10⁴ ∼ 7.5 x 10⁴. The effect of mass flux and heat flux was dominant factor in the supercritical pressure heat transfer while the effect of pressure was negligible. The Bishop's correlation predicted the test result most closely and Bae and Kim's recent correlation was the next. The heat transfer deterioration occurred when GR)b/Re_b^2.7 > 2.0 x 10^-5. As soon as the heat transfer was deteriorated, it entered a new regime and did not recover the normal heat transfer nevertheless Gr_b/Re_b^2.7 reduced below 2.0 x 10^-5. It may mean that the correlation must be developed for the normal and deterioration regime separately

[en] The SCWR(Super-Critical Water cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project which aims at the development of new reactors with enhanced economy and safety. The SCWR is considered to be a feasible concept of new nuclear power plant if the existing technologies developed in fossil fuel fired plant and LWR technologies together with additional research on several disciplines such as materials, water chemistry and safety. As KAERI takes part in the GIF(Generation IV Forum) for the Gen-IV project, domestic concerns about the SCWR have been recently increased. In order to establish a foundation for the development of SCWR, efforts should be concentrated on the conceptual design of systems and the associated key experiments as well. Heat transfer experiments, among others, under supercritical condition are required for the proper prediction of thermal hydraulic phenomena, which are essential for the thermal hydraulic designs of reactor core. Nevertheless, the experiments have not been performed in Korea yet. This report deals with fundamental surveys on the heat transfer experiments under supercritical conditions, which are required for the understanding of heat transfer characteristics for the thermal hydraulic designs of supercritical reactor core. Investigations on the physical properties of water and CO₂ showed that the physical properties such as density, specific heat, viscosity and thermal conductivity are significantly changed near the pseudo-critical points. The state of the art on the heat transfer characteristics in relation with heat transfer deterioration and heat transfer coefficient is briefly described. In addition, previous experiments with supercritical water as well as supercritical CO₂ and Freon used for an alternating fluid are presented

[en] The hydraulic diameter of a subchannel in a core concept developed at KAERI is 6.5 mm. The sub-channel is much smaller than that of the conventional PWR, and naturally a helical wire was considered as one of the candidates for a spacer. For simplicity the subchannel is simulated by a commercially available Inconel 625 tube of 6.32 mm ID with a helically-coiled spring steel wire insert of 1.3 mm OD. The medium is CO₂. The test pressures are 7.75 and 8.12 MPa corresponding to 1.05 and 1.1 times the critical pressure of CO₂, respectively. The mass flux and heat flux, which were in the range of 400 ∼ 1200 kg/m²s and 30 ∼ 90 kW/m² respectively, were varied at a given system pressure. The corresponding Reynolds numbers at the inlet spans between 2.5 x 10⁴ and 7.5 x 10⁴. It was observed that the heat transfer was enhanced by almost twice in most of the tested enthalpy range except for in the the region far from the pseudocritical point. The test results revealed that the wire effect was sustained in the downstream up to 40-60 times the wire diameter. The temperature decreased in the first half of the span between contact points and it increased in the second half of the span

[en] An auto-door hinge, which is one of the automatic door-closing apparatuses, has been widely used to prevent fire propagations in living or commercial buildings. The auto-door hinge consists of a spring to accumulate power for closing a door and an oil damper to control door-closing velocity. To predict and optimize the temporal door behavior during the door-closing period, the auto-door closing system was modeled as a second order-damping system. And a damping coefficient of the oil damper was also theoretically modeled by analyzing Newtonian, incompressible, viscous flow through an oil passage between a oil control rod and a oil piston body. The temporal door behaviors during the door-closing period were predicted with respect to the gap distance of the oil passage, oil viscosity, and pre-compressing of the spring. Temporal door behavior measurement method using an encoder system was also developed to validate the modelling on the oil damping system. As using the developed test apparatus, the temporal door position, velocity, and rotational torque were measured, and the modelling method was evaluated

[en] Spent nuclear fuel contains useful nuclides as valuable resource materials for energy, heat and catalyst. High-level wastes (HLW) are expected to be generated from the R and D activities and reuse processes. It is necessary to develop vitrification or advanced solidification technologies for the safe long-term management of high level wastes. As a first step to establish HLW vitrification technology, characterization of HLWs that would arise at KAERI site, glass melting experiments with a lab-scale high frequency induction melter, and fabrication and property evaluation of base-glass made of used HEPA filter media and additives were performed. Basic study on the fabrication and characterization of candidate ceramic waste form (Synroc) was also carried out. These HLW solidification technologies would be directly useful for carrying out the R and Ds on the nuclear fuel cycle and waste management. (author). 70 refs., 29 tabs., 35 figs

[en] Several experiments have been performed in the VESTA (Verification of Ex-vessel corium STAbilization) facility using a prototypic melt and APR1400 ICI penetration test specimens. In the previous experiments, zirconia melt about 2500 .deg. C was generated in a cold crucible by induction heating and interacted with the test specimen at high pressure condition to simulate the severe accident environment in a reactor vessel. In order to find the tube ejection criteria, we need to check the experimental conditions and find proper boundary conditions to induce tube ejection in the test facility. The objective of the present paper is to investigate the gap change (i.e., to identify the contact or non-contact status) between the tube and the reactor vessel hole, and to estimate the tube ejection for the APR1400 ICI penetrations according to various boundary conditions such as the constraint to fix the test specimen, external reactor vessel cooling, and melt penetration inside the tube. It was found that the structural constraint of the reactor vessel hole not to expand outward and large temperature differences between the penetration tube and the reactor vessel reduces the tube ejection probability. That is, when the thermal expansion of the reactor vessel is suppressed and the temperature difference becomes large, the tube touches the reactor vessel hole early, and thus the tube ejection would not occur. Moreover, the melt flow inside the tube turned out have the greatest effect on the early gap contact because it leads to large temperature difference

[en] Tungsten carbide (WC) and dense WC-5 vol.% Co materials, with grain size of ∼1 μm, were synthesized by high-frequency induction heated combustion synthesis (HFIHCS) method in one step from elemental powders of W, C, and cobalt (Co) within several minutes. Simultaneous combustion synthesis and densification were accomplished under the combined effects of an induced current and mechanical pressure. In the absence of cobalt additive, WC can be formed, but its relative density was low (about 70%) under simultaneous application of a 60 MPa pressure and the induced current. However, in the presence of 5 vol.% cobalt, the density increased to 98.5% under the same experimental conditions. The percentages of the total volume shrinkage occurring before and during the synthesis reaction of WC-5 vol.% Co were 7 and 57%, respectively. The fracture toughness and hardness values obtained of WC-5 vol.% Co were 7 MPa m^1/2 and 2100 kg/mm², respectively