[en] This thesis applies new information techniques, artificial neural networks, (ANNs) and fuzzy theory, to the investigation of the critical heat flux (CHF) phenomenon for water flow in vertical round tubes. The work performed are (a) classification and prediction of CHF based on fuzzy clustering and ANN, (b) prediction and parametric trends analysis of CHF using ANN with the introduction of dimensionless parameters, and (c) detection of CHF occurrence using fuzzy rule and spatiotemporal neural network (STN). Fuzzy clustering and ANN are used for classification and prediction of the CHF using primary system parameters. The fuzzy clustering classifies the experimental CHF data into a few data clusters (data groups) according to the data characteristics. After classification of the experimental data, the characteristics of the resulted clusters are discussed with emphasis on the distribution of the experimental conditions and physical mechanisms. The CHF data in each group are trained in an artificial neural network to predict the CHF. The artificial neural network adjusts the weight so as to minimize the prediction error within the corresponding cluster. Application of the proposed method to the KAIST CHF data bank shows good prediction capability of the CHF, better than other existing methods. Parametric trends of the CHF are analyzed by applying artificial neural networks to a CHF data base for water flow in uniformly heated vertical round tubes. The analyses are performed from three viewpoints, i.e., for fixed inlet conditions, for fixed exit conditions, and based on local conditions hypothesis. In order to remove the necessity of data classification, Katto and Groeneveld et al.'s dimensionless parameters are introduced in training the ANNs with the experimental CHF data. The trained ANNs predict the CHF better than any other conventional correlations, showing RMS error of 8.9%, 13.1%, and 19.3% for fixed inlet conditions, for fixed exit conditions, and for local conditions hypothesis, respectively. The parametric trends of the CHF obtained from those trained ANNs show a general agreement with previous understanding. In addition, this study provides more comprehensive information and indicates interesting points for the effects of tube diameter, heated length, and mass flux. Finally, CHF detection methods based on simple fuzzy rule and STN are developed and applied to the low-pressure and low-flow CHF and pool boiling CHF. The CHF occurrences are well detected using two methods

[en] A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). The ATLAS is a half-height and 1/288-volume scaled test facility with respect to the APR1400. The fluid system of the ATLAS consists of a primary system, a secondary system, a safety injection system, a break simulating system, a containment simulating system, and auxiliary systems. The primary system includes a reactor vessel, two hot legs, four cold legs, a pressurizer, four reactor coolant pumps, and two steam generators. The secondary system of the ATLAS is simplified to be of a circulating looptype. Most of the safety injection features of the APR1400 and the OPR1000 are incorporated into the safety injection system of the ATLAS. In the ATLAS test facility, about 1300 instrumentations are installed to precisely investigate the thermal-hydraulic behavior in simulation of the various test scenarios. This report describes the scaling methodology, the geometric data of the individual component, and the specification and the location of the instrumentations which are specific to the simulation of 50% DVI line break accident of the APR1400 for supporting the 50^th OECD/NEA International Standard Problem Exercise (ISP-50)

[en] An experimental study of bubble distribution at various flow regimes has been carried out in air-water two-phase flows flowing upward in a 40mm diameter acrylic tube under atmospheric pressure. To obtain local void fraction, U-shaped optical probe is used. The probe signal is amplified by electric circuit. The amplified signal is sampled uniformly using the analog to digital converter and stored and processed in IBM-PC/AT computer. The probe was very sensitive to the existence of air and water phase and can be used to obtain not only local void fraction but also interesting parameters

[en] KAERI has performed an experimental study of water Critical Heat Flux (CHF) under zero flow conditions with a non-uniformly heated 3 by 3 rod bundle. Experimental conditions are in the range of a system pressure from 0.5 to 15.0 MPa and inlet water subcooling enthalpies from 67.5 to 351.5 kJ/kg. The test section used in the present experiments consisted of a vertical flow channel, upper and lower plenums, and a non-uniformly heated 3 by 3 rod bundle. The experimental results show that the CHFs in low-pressure conditions are somewhat scattered within a narrow range. As the system pressure increases, however, the CHFs show a consistent parametric trend. The CHFs occur in the upper region of the heated section, but the vertical distances of the detected CHFs from the bottom of the heated section are reduced as the system pressure increases. Even though the effects of the inlet water subcooling enthalpies and system pressure in the flooding CHF are relatively smaller than those of the flow boiling CHF, the CHF increases by increasing the inlet water subcooling enthalpies. Several existing correlations for the countercurrent flooding CHF based on Wallis's flooding correlation and Kutateladze's criterion for the onset of flooding are compared with the CHF data obtained in the present experiments to examine the applicability of the correlations

[en] The injected coolant decreases temperature of the heated fuel rods for 40 - 200 second (reflood). After core vessel is filled by coolant, it is long-term cooling. During the single-phase steam flow in the early phase of the reflood, the cladding temperature may increase and have a peak value due to low heat transfer from the fuel to the steam. The increased temperature can make a ballooned fuel rods. As a result, the flow passage area of sub-channel is reduced and it leads the redistribution of flow and heat transfer in sub-channels. The flow rate through the sub-channel between ballooned fuel rods is decreased while the flow rate through the sub-channel between intact fuel rods is increased. The reduction of flow reduces the capacity of coolability and ballooned fuel rods have higher temperature than non-deformed fuel rods. If a LBLOCA condition and ballooned fuel rods are occurred, the effect of reduced flow passage on the convective heat transfer by single-phase steam flow is important phenomena to analyze the safety of a reactor. During the LOCA condition, accumulation of fuel debris in the ballooned region of the bust cladding, which resulted from fuel fragments slumping from upper regions, can be occurred.. This fuel relocation makes different thermal hydraulic behavior. The present experiments were performed in various Reynolds numbers (about 2600 - 13000) and Heater power (0.14 - 1.12 kW/m) to investigate the effect of the Fuel-relocation on heat transfer phenomena by single-phase steam flow. The experiments were performed in three rod bundles in KAERI reflood ATHER test facility. One is a non-deformed 6 x 6 rod bundle, which consists of 36 non-deformed heater rods. Experimental study of heat transfer phenomena by single-phase steam flow using three type heater bundles (intact bundle, ballooned bundle, fuel-relocated bundle) were performed to investigate the effect of fuel-relocation on single-phase steam flow. Fuel-relocated bundle showed its own characteristics

[en] An experimental study on Critical Heat Flux(CHF) has been performed for water flow in a uniformly heated vertical 3 by 3 rod bundle under low flow and a wide range of pressure conditions. The objective of this study is to investigate the parametric trends of CHF with 3 by 3 rod bundle test section where three unheated rods exist. The general trends of the CHF are coincident with previous understandings. At low flow and system pressure above 3 Mpa, some critical qualities are larger than 1.0 due to counter-current flow in test sections. Since there is a supply of water to the heated section from unheated section, the maximum CHFs at system pressure between 2 and 4 Mpa are not shown

[en] The experimental study of water CHF (Critical Heat Flux) under zero flow conditions were carried out in an annulus flow channel with uniformly and non-uniformly heated sections over a pressure range of 0.52 to 14.96 MPa. A comparison of the present data with the existing flooding CHF correlations shows that the predicted values by the existing flooding CHF correlations give considerably lower values than the present data. When the correction terms with the density ratio and the effect of the heat flux distribution proposed in the present work are used with the CHF correlation based on the Wallis flooding correlation, it predicts the measured flooding CHF within an RMS error of 9.0 %

[en] A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). The ATLAS has the same two-loop features as the APR1400 and is designed according to the well-known scaling method suggested by Ishii and Kataoka to simulate the various test scenarios as realistically as possible. It is a half-height and 1/288-volume scaled test facility with respect to the APR1400. The fluid system of the ATLAS consists of a primary system, a secondary system, a safety injection system, a break simulating system, a containment simulating system, and auxiliary systems. The primary system includes a reactor vessel, two hot legs, four cold legs, a pressurizer, four reactor coolant pumps, and two steam generators. The secondary system of the ATLAS is simplified to be of a circulating loop-type. Most of the safety injection features of the APR1400 and the OPR1000 are incorporated into the safety injection system of the ATLAS. In the ATLAS test facility, about 1300 instrumentations are installed to precisely investigate the thermal-hydraulic behavior in simulation of the various test scenarios. This report describes the scaling methodology, the geometric data of the individual component, and the specification and the location of the instrumentations in detail

[en] SPACE, which is a safety analysis code for nuclear power plants, has been developed to analyze the multidimensional, two-component and three-field flow. This code can be applied to safety analysis for approval which is thermal-hydraulic analysis to support the nuclear power station design, establishment of accident ease strategy, development of operating guide line, experiment plan and analysis. To do so, SPACE code has 12 wall heat transfer mode and the corresponding models and correlations to deal with the physical heat transfer phenomenon in wall surface. In this report, the physical correlation models regarding the wall heat transfer are explained and their performance is assessed against several SET

[en] KAERI has performed an experimental study of water critical heat flux (CHF) under zero flow conditions with a non-uniformly heated 3 by 3 rod bundle. Experimental conditions are in the range of a system pressure from 0.5 to 15.0 MPa and inlet water subcooling enthalpies from 67.5 to 351.5 kJ/kg. The test section used in the present experiments consisted of a vertical flow channel, upper and lower plenums, and a non-uniformly heated 3 by 3 rod bundle. The experimental results show that the CHFs in low-pressure conditions are somewhat scattered within a narrow range. As the system pressure increases, however, the CHFs show a consistent parametric trend. The CHFs occur in the upper region of the heated section, but the vertical distances of the detected CHFs from the bottom of the heated section are reduced as the system pressure increases. Even though the effects of the inlet water subcooling enthalpies and system pressure in the flooding CHF are relatively smaller than those of the flow boiling CHF, the CHF increases by increasing the inlet water subcooling enthalpies. Several existing correlations for the countercurrent flooding CHF based on Wallis's flooding correlation and Kutateladze's criterion for the onset of flooding are compared with the CHF data obtained in the present experiments to examine the applicability of the correlations