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[en] In this study, mainly two experimental works have been done. The first one is the non-heating simulation of pool boiling and CHF using the air-blowing drilled flat plate and the related analytical study on the formation of macrolayer. And the second is the CHF experiments for the in-vessel retention strategy of severe accident. First, the non-heating experimental method was proposed and confirmed by the experiments and analytical study. The proposed non-heating experimental methodology is based on the macrolayer dryout model: by measuring the initial macrolayer thickness and bubble departure frequency from the non-heating experiments, the onset of CHF can be predicted based on the macrolayer thickness. Through the non-heating experiments, the void fraction, the macrolayer thickness and the bubble departure frequency were measured. The macrolayer thickness measured by the conduction probe was generally well agreed with the previous steam-water experiments in its trend and magnitude. And, the predicted CHF values by the proposed methodology were generally agreed well with the boiling CHF data for horizontal, flat plate. On the other hand of experimental works, an analytical study on formation of the macrolayer and its relation with the CHF was conducted. Through this analytical study, we found that the lateral coalescence of the growing bubble was the underlying mechanism of the macrolayer formation. Through this experimental and analytical study, the feasibility of the non-heating simulation was confirmed. Second, the CHF on the reactor vessel outer wall was measured using 2-dimensional slice test section. Because there is few CHF data for the mass flux less than 200 kg/m2sec, in this experimental study, the mass flux effect on the CHF was mainly studied. The radius, the width of heater and the channel gap were 2.5 m, 10 cm and 15 cm relatively. The CHF data under the inlet subcooling of 2∼14 K and the mass flux 6∼210 kg/m2sec had been acquired and a CHF correlation had been proposed. The measured CHF value was generally smaller than that of ULPU. However, the general trend of the CHF according to mass flux was similar with that of ULPU. And the experimental CHF data were compared with predicted values by the SULTAN correlation. The SULTAN correlation predicted well this study's data for high mass flux (150 kg/m2sec ∼). But, it couldn't predict well under low mass flux condition. The developed local-condition-based correlation showed good prediction capability for low quality (-0.01∼0.44) and low mass flux less than 210 kg/m2sec with RMS error of 1.9%
[en] During the severe accident which can lead to core melt, particle bed may be formed in vessel or on the cavity floor by quenching of corium. If the heat removal from the particle bed is insufficient, reactor vessel or reactor cavity floor may be attacked by re-melted corium. Therefore, the prediction of DHF with high accuracy is important for the design and safety analysis for severe accident. The DHF is affected by many parameters such as, geometry of particle bed, system pressure, coolant properties, bottom flow, etc. Until now, considerable experimental or theoretical studies have been performed. However, the useful DHF correlation for various kinds of conditions, specially low pressure and low bottom flow conditions, has not been recommended. In this regard, the existing correlations have been assessed and the new correlation that can predict DHF under low pressure and low bottom flow conditions, has been developed. In the other hand, during the quenching process of corium, violent fuel-coolant interaction, named steam explosion, can be occurred. In this process, film boiling heat transfer will occur by the great temperature difference between corium and coolant. Therefore, to exactly estimate the heat transfer rate to the coolant, film boiling heat transfer coefficient should be identified, especially on spheres. Especially, in severe accident condition, corium and water may interact with each other under single- and two-phase conditions with some relative velocity between them. In this regard, a series of correlations of film boiling heat transfer on spheres under single- and two-phase flow conditions have been developed based on heat transfer correlations without phase change. The correlations have been developed for the conditions of saturated single-phase flow, subcooled single-phase flow, upward two-phase flow and downward two-phase flow conditions, respectively. They predict well the film boiling heat transfer coefficients within ±20 % of error bounds
[en] There are three major ways to generate hydrogen from water. These are pure electrolysis, pure thermochemical methods and hybrid methods. Pure electrolysis splits water molecules into hydrogen and oxygen using electrical energy. A pure thermochemical method is a series of chemical reactions to split water at high temperature, but much lower temperatures than that of direct dissociation of water (4000 .deg. C). The hybrid cycle is a combination of electrolysis and thermochemical cycle. The hybrid sulfur cycle (often called the Westinghouse Cycle) has two steps for decomposing water into hydrogen and oxygen. Hydrogen and sulfuric acid are produced by electrolysis of a sulfur dioxide and water mixture at low temperature. The sulfuric acid is decomposed into sulfur trioxide and steam and the sulfur trioxide is decomposed into sulfur dioxide and oxygen at high temperature (∼1100 K). The sulfur dioxide is supplied to the electrolyzer. In this study, we develop detailed flow sheets for the hybrid sulfur cycle and estimate the overall efficiency of the hydrogen production process
[en] In this study, the feasibility of natural circulation was evaluated for the reference plant AHR400 (Advanced Heating Reactor 400MWth). AHR400 is a pool-type desalination-dedicated nuclear reactor. As a consequence, AHR400 has low operating pressure and temperature which provides large safety margin. Removal of the reactor coolant pump from the AHR400 will enforce integrity of the reactor vessel and passive safety feature. Therefore, the study also tried to find out optimized primary loop design to achieve total natural circulation of the coolant. Natural circulation capacity of the primary loop of the desalination dedicated nuclear reactor AHR400 was evaluated. It was concluded that to remove RCP from the AHR400 and operates the reactor only by natural circulation of the coolant is impossible. Decreased core power as half make removal of RCP possible with 15m central height difference between the core and IHXs. Furthermore, validation and modification of pressure loss coefficients by small-scaled natural circulation experiment at a pool-type reactor would provide more accurate results
[en] The dry cooling systems commonly use extended surface, called fin, to increase surface area and heat transfer performance. The finned surfaces ideally have zero thermal resistance and show uniform temperature distribution. Direct contact heat transfer is getting interests as innovative solution in the dry cooling system. Hot fluid flows into the air and heat transfer occurs without physical barrier between the fluid and air, thus the heat transfer performance is high. For getting large surface area in the direct contact heat transfer system, the hot fluid is emitted as small droplet shape or thin falling film. UCLA published the Direct Contact Liquid-on-String Heat Exchanger (DILSHE) that the hot fluid flows along strings, and is cooled by air. Drexel published Spray Freezing of Recirculating PCM system that the hot fluid is sprayed as droplet shape into air and the cooled fluid returns to the pool. In KAIST, dry cooled waste heat removal system that applies the direct contact heat transfer technology has been being designed. This study conducted analysis for direct contact heat transfer on the film along vertical straight string. Physical modeling on the film, heat transfer analysis and pressure loss, and cooling tower analysis were conducted. When the string radius is larger than 1 mm, the film along the string doesn't break and stable film forms. As the string radius increases, the required cooling tower height is increased but the increase is very small.
[en] The Korean large pressurized water reactor APR1400 applied passive flow controller for the emergency core cooling system to correspond to the increase of concerns about passive safety. The passive flow controller in the APR1400, which is called fluidic device, controls the flow rate of emergency core cooling (ECC) water from the safety injection tank (SIT). The vortex chamber in the fluidic device controls flow rate. The vortex builds up high flow resistance through the fluidic device and therefore, the discharge flow rate decreases. The performance of the fluidic device was evaluated by experiments with the valve performance evaluation rig (VAPER) facility in the KAERI. The discharge mass flow rate from the SIT over time was demonstrated as expected. However, the experiment did not focus on the nitrogen entrainment phenomena which has been issued recently. The concept of equilibrium water level was brought to analyze the nitrogen entrainment phenomena and to figure out the criteria of the nitrogen entrainment. Deduced equilibrium water level is 19.5 cm for the initial condition of 10 bar. The level increases through the initial pressure increases. The longer stand pipe which can provide enough water level difference between stand pipe and SIT will decrease the amount of entrained nitrogen. The large, short and simple-shaped control pipe which lowers the pressure loss at the pipe will decrease the amount of entrained nitrogen. The results can be used to design a SIT with no nitrogen entrainment to the primary loop of the nuclear reactor.
[en] In this study a new desalination system using the waste heat of nuclear power plant or solar energy system is under consideration. An experiment study will performed to evaluate the performance of the system. In this paper, the experimental design of the system and steady-state and transient analysis, using MATLAB and MARS code respectively, will be presented. A new desalination system using the waste heat from nuclear power plants or solar energy was proposed and the design of an experimental facility to evaluate the system's performance was presented. The analytical steady-state analysis using MATLAB and transient MARS calculations have been performed. An Experiment will be performed to verify the model. Moreover a feasibility study for the proposed design and a comparison with the conventional desalination system will be performed.
[en] The APR+ (Advanced Power Reactor Plus), which is currently being developed in South Korea, has adopted a Passive Auxiliary Feedwater System (PAFS), which removes the decay heat through the steam generator replacing the conventional active auxiliary feedwater system. The steam condenses under high pressure (7.4 MPa, 290 C) in 8.4m long horizontal tube, with 3 degrees inclination, with inner diameter of 44.8mm and wall thickness of 3mm. the PCHX consists of 4 bundles, each bundle has 60 condensation tubes. A separate effect test facility using the PASCAL (PAFS Condensing Heat Removal Assessment Loop) experimental facility to verify the cooling and operational performance of the PAFS had been conducted. Evaluation of the one-dimensional safety analysis code MARS (Multidimensional Analysis of Reactor Safety) against the experimental data showed that the MARS code under predict the local heat transfer coefficient. In this study a scaling criteria have been developed to simulate a high pressure steam in horizontal tube condensation by a reduced pressure model using simulant fluid. The criteria have been developed based on the Buckingham Pi theorem and the compensated distortion technique used in Ahmad's paper. The developed scaling criteria have been verified by comparing different prototypes and models using three different condensation models. The prototype-equivalent and the prototype localheat transfer coefficients showed good agreement with each other, based on the performed statistical analysis. As a future work, in order to verify the current scaling criteria, dimensional analysis based on the non- dimensionalization of the governing equations (continuity, momentum and energy) and the appropriate boundary conditions of each phase will be performed.
[en] Two-phase closed thermosyphon (TPCT) is vertically oriented wickless heat pipe that has working fluid in the interior. The TPCT transports a large amount of heat from evaporator to condenser by phase change of working fluid, and the working fluid passively returns to evaporator by gravity. Due to these advantages of the TPCT, the TPCT is considered as method of PRHR (Passive Residual Heat Removal) system in nuclear system. Parametric studies have done to investigate the heat transfer characteristics of the TPCT. Different working fluids such as water, ethanol, methanol and acetone were used at various filling ratios and at different operating temperatures to find maximum heat transport capabilities of TPCT. Effect of heat transfer rate, filling ratio and aspect ratio were investigated. Inclined angle effect was investigated at several filling ratios and working fluids. This study is interested in silicon oil effect on the TPCT. To carry out the experiment, experimental apparatus is designed and manufactured. In design process, the TPCT operation limit is considered This study is interested in silicon oil effect on the TPCT. Experiments were carried out at three oil weight percent with three input power. Effect of oil on the TPCT is evaluated by inner wall temperature distribution and thermal resistance. In this study, silicon oil effect on TPCT was investigated. The TPCT was operated with several oil weight percent and input power. From experiment, overall, the silicon oil reduced evaporator thermal performance, but enhanced condenser thermal performance. However, the TPCT total thermal performance was reduced by 100 c St silicon oil
[en] Critical heat flux (CHF) of the external reactor vessel wall is a safety limit that indicate the integrity of the reactor vessel during the situation. Many research conducted CHF experiments in the IVR-ERVC conditions. However, the flow velocity field which is an important factor in the CHF mechanism were not studied enough in the IVR-ERVC situations. In this study, flow measurements including velocity vector field and the liquid velocity in the IVR-ERVC conditions were studied. The air-water two phase flow loop simulating IVRERVC conditions was set up and liquid velocity field was measured by LIF/PIV technique in this study. The experiment was conducted with and without air injection conditions. For the air-water flow experiment, liquid velocity at the outside of two phase boundary layer became higher and the two phase boundary layer thickness became smaller when the mass flux increases. The velocity data obtained in this study are expected to improve the CHF correlation in the IVR-ERVC situations.