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The flow characteristics at the shell side of a prototype intermediate heat exchanger (p-IHX) in a prototype generation-IV sodium-cooled fast reactor (PGSFR) were investigated experimentally in this study. The shell side of the p-IHX consisted of an inlet window, tube bundles, grid plates, and an exit flow channel. Two different test facilities were newly constructed to identify the pressure drop characteristics due to the complicated configuration with minimized scaling distortion. An intermediate heat exchanger test loop for PGSFR (iHELP) was constructed to characterize the flow resistance of the tube bundle regions with the grid plates and inlet window. The iHELP test section with a slab-shaped cross section was manufactured with a 1/29.6 vol scale ratio by preserving the height of the p-IHX. The hydraulic diameter at the tube bundle regions, the porosity of the grid plates, and the configuration of the tube bundles were conserved. In addition, the Reynolds number was preserved, which determined the flow rate conditions for the iHELP experiments. To investigate the pressure drop characteristics of the exit flow channel (EC), an intermediate heat exchanger test loop for the exit flow channel (IEC) was built with a 1/5 linearly reduced length scale by preserving the geometry of the flow path based on the EC of the p-IHX. The Euler number was conserved with a 1/8 Reynolds number ratio corresponding to a 1/1 flow velocity ratio to the p-IHX for dynamic similarity. By using each test facility the experimental pressure drop data satisfying the experimental uncertainty requirements were obtained separately for the bundle and exit channel regions. The pressure drop at the shell side of the p-IHX under a wide range of flow rate conditions was determined by using the pressure drop results from the iHELP and the IEC test facilities, which would be useful in improving and validating the pressure drop correlations in the IHX design computational code (SHXSA).

The supercritical CO2 Brayton cycle has received attention as a next generation power conversion system. However, technical challenges that relate to the design and operation of the system remain to be overcome. In particular, achieving turbomachinery operation with design points is imperative to demonstrate the feasibility of the technology. This study aims to provide the current research status of the supercritical CO2 power generation system in KAERI, with a focus on the detailed design of the system and operation of the compressor. In this study, cycle design (a simple recuperated cycle layout) of the system was developed using an in-house code for the optimization of a pilot plant of a 500 kW supercritical CO2 power generation system for waste heat recovery. A thermal efficiency of 16.5% of the proposed cycle was achieved. In addition, the detailed design of key components, such as turbomachines and heat exchangers, was developed. Furthermore, the first-step configuration, a compressor performance test loop, of the supercritical CO2 power generation system pilot plant was accomplished. The compressor was designed and manufactured as a hermetic system for the supercritical CO2 power generation system. A compressor performance test was conducted to verify the target performance of the compressor. The operation of the compressor at the design point was achieved during the test; the compressor inlet was 33 °C at 7.68 MPa, and the compressor rotational speed was 34200 rpm; its efficiency was 83.7%, and the pressure ratio was 1.756.

Flow-induced vibration characteristics of a U-tube bundle were experimentally investigated in air–water two-phase flow. The test section was equipped with 39 U-tubes, simulating the innermost region of an actual steam generator. The U-tubes were made of Inconel 690 with a diameter of 19.05 mm. The horizontal region of the U-tubes had a rotated square array with a pitch of 31.11 mm and a p/d of 1.633. The U-tubes and supporting structures have almost the same prototypical geometries. Vibration responses of six U-tubes were measured with ten 3-axis accelerometers. Two sets of experiments were performed to investigate an onset of fluid-elastic instability, damping ratio, and hydrodynamic mass of the U-tubes. The experiments were performed for a void fraction of 70–95%. The instability constant (K) of the Connors’ equation for the present U-tube bundle was evaluated to be in the range of 6.5–10.5.

A nucleate boiling limitation model which is applicable to the heat transfer prediction in the nucleate boiling region and the CHF was proposed for a pool boiling. The present model was developed based on the direct observations of the physical boiling phenomena. The predicted boiling curves for the nucleate boiling region agree well with both the vertical and the horizontal surface data for all the contact angles. The predicted CHF for the vertical surface also agrees well with the experimental data, but the present model underpredicts the CHF by about 30% for the horizontal surface data.

A new experimental attempt was made to simultaneously observe the dynamic behaviors of bubbles and dry spots in the vicinity of boiling surface. Also, the two-dimensional bubble structures were obtained separately. From the visualization results, the formation of bubbles and dry spots occurs simultaneously. At critical heat flux (CHF), the surface rewetting is repeated by the local nucleate boiling around the large vapor film. At just after CHF, nucleate boiling at the locally wetted region is extinguished, resulting in the dryout of the whole heater surface. Therefore, we conclude that CHF is initiated from the locally limited nucleate boiling activity rather than any hydrodynamic instability.

Direct-contact condensation experiments on atmospheric steam and steam/air mixture on subcooled water flowing co-currently in nearly-horizontal channels are carried out and the local heat transfer coefficients are obtained. Inlet air mass fraction in the mixture is varied by up to 50%. Based on previous and present experimental data, a direct-contact condensation database is constructed. The horizontally stratified condensation model of RELAP5/MOD3.2 overpredicts both co-current and counter-current experimental data. The correlation proposed by Kim predicts the database relatively well compared with that of RELAP5/MOD3.2. In the presence of noncondensable gases, RELAP5/MOD3.2 overpredicts the database with percentage errors of 55.1 and 61.1% for pipe inclination angles of θ=2.1° and θ=5.0°, respectively. The UCB correction factor is modified to consider the effects of noncondensable gases on heat transfer coefficients. When Kim's correlation is substituted with the Dittus–Boelter type correlation in RELAP5/MOD3.2 and a modified correction factor is used, the prediction errors are greatly reduced to 20.7 and 28.8% for inclination angles of θ=2.1° and θ=5.0°, respectively.

Korea Atomic Energy Research Institute (KAERI) has designed and constructed a test facility for reactor coolant pumps (RCPs). The RCP Test Facility (RCPTF) has the capability to test a RCP under the operation condition of an Advanced Power Reactor 1400 MW (APR1400). The design values of the facility are 17.2 MPa, 343 °C, 11.7 m3/s, and 13 MW in maximum pressure, temperature, flow rate, and electrical power, respectively. In the facility, it is possible to perform a type test for a newly-developed RCP as well as a production test for a RCP before its installation in a nuclear power plant. For the production test, H-Q curves under the cold and hot conditions are acquired. For the type test, various transient tests are additionally performed including four types of seal transient tests, a thrust bearing transient test, a cost down test, and so on. To acquire H-Q curves of a RCP, the flow rate should be controlled by varying the flow resistance in the test loop. The RCPTF uses a Variable Restriction Orifice (VRO) whose flow area can be controlled by moving the two orifice plates installed in-parallel. The need for flow control valves and bypass lines was eliminated using the VRO such that the flow disturbance was minimized. The flow rate in the main loop of the RCPTF is measured by a standard venture flow meter. The flow rate in the RCPTF is very high and thus the venture flow meter could not be calibrated in the entire range of Reynolds number corresponding to the operating condition in the APR1400. The calibration was conducted at the Colorado Experiment Engineering Station Inc. (CEESI) in the USA where natural gas is used for a working fluid. If a discharge coefficient calibrated with the gas is applied in the test results performed using the water as a working fluid, a discrepancy can occur due to the static hole error. Therefore, the static hole error was compensated in the test results and the result shows the improvement. The effect of the temperature on the pressure pulsation amplitude was also evaluated. During a cold performance test and heat-up phase to the condition of a hot performance test, an abnormal increase in the pressure pulsation amplitude was observed near the specific temperature range. This is acoustic resonance phenomena that occur when a blade passing frequency of the RCP is proportional to the harmonic resonance frequency of the RCPTF.

An experimental study has been carried out to investigate an I-type sparger performance in view of pressure oscillation and thermal mixing in a pool. Its pitch-to-hole diameter, P/D, varies from 2 to 5. The test conditions are restricted to the condensation oscillation regime. In the present study, two different hole patterns, staggered and parallel types, are employed under various test conditions. The amplitude of the pressure pulse shows a peak for pool temperatures of 45∼85°C, which depends on P/D and the steam mass flux. The effect of hole pattern on the pressure load is smaller than that ofP/D. The dominant frequency increases with the subcooling temperature of pool water and P/D. A correlation for the dominant frequency is proposed in terms of the pitch-to-hole diameter ratio and other dimensionless thermal hydraulic parameters.

The experimental study of water CHF (critical heat flux) under zero flow conditions has been carried out in an annulus flow channel with uniformly and non-uniformly heated sections over a pressure range of 0.52–14.96 MPa. In the present boiling system, the CHFs occur in the upper region of the heated section, in contrast to the results in the experiments for boiling tubes conducted by several investigators. The general trend of the CHF with pressure is that the CHF increases up to a medium pressure of about 6–8 MPa and decreases as the pressure is further increased. A comparison of the present data with the existing flooding CHF correlations shows that the correlations depend greatly on the effect of the heat flux distribution. 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%.

Critical heat flux (CHF) experiments have been carried out in a wide range of pressure for an internally heated vertical annulus. The experimental conditions covered a range of pressure from 0.57 to 15.01 MPa, mass fluxes of 0 kg m−2 s−1 and from 200 to 650 kg m−2 s−1, and inlet subcoolings from 85 to 413 kJ kg−1. Most of the CHFs were identified to the dryout of the liquid film in the annular-mist flow. For the mass fluxes of 550 and 650 kg m−2 s−1, the CHFs had a maximum value at a pressure of 2–3 MPa, and the pressure at the maximum CHF values had a trend moving toward the pressure at the peak value of pool boiling CHF as the mass flux decreased. The CHF data under a zero mass flux condition indicate that both the effects of pressure and inlet subcooling on the CHF were smaller, compared with those for the CHF with a net water upflow. The Doerffer correlation using the 1995 CHF look-up table and the Bowring correlation show a good prediction capability for the present CHF data.

The hydraulic characteristics in subchannels of axially finned rod bundles installed in the Korea Multipurpose Research Reactor (KMRR) were measured using one-component laser-Doppler velocimetry (LDV). Pressure drops for each component, time mean axial velocity, and axial turbulent intensity were measured. Then the friction factors in rod bundles were estimated from the measured pressure drops. The turbulent crossflow mixing rate between neighboring subchannels was evaluated from the measured data. The results show that (1) the friction factors for axially finned rod bundles are less than those given by Moody's correlation for smooth tubes, and (2) as the flow develops, the turbulent crossflow mixing rate between neighboring subchannels decreases and, in the developed region, the values level out.

Fluid-elastic instability characteristics of a U-tube bundle were experimentally investigated in air-water two-phase flow. A total of 39 U-tubes were arranged in a rotated square array with a pitch-to-diameter ratio of 1.633. Vibration responses of four U-tubes were measured with three-axis accelerometers. Two sets of experiments were performed to investigate the onset of fluid-elastic instability, and the damping and hydrodynamic mass of the U-tube. The experiments were performed for a void fraction of 70–95%. Fluid-elastic instability was clearly observed in an out-of-plane mode vibration. The effect of a primary side flow on the vibration of U-tube was investigated separately. The damping ratio of the present U-tube was higher than the damping ratio of the cantilever tubes in the literature. The hydrodynamic mass of the U-tube was generally in accordance with the hydrodynamic mass of the cantilever tubes in the literature. The instability constant (K) of the Connors equation was assessed with a simplified effective gap velocity, and the fluid-elastic instability constant was 8.5.

Fluid-elastic instability characteristics in an air-water two-phase cross-flow have been experimentally investigated using two different arrangements of cantilevered straight tube bundles. Rotated triangular array tube bundle is for the supplementary test of the existing work, and normal square array tube bundle is for the investigation of fluid-elastic instability in higher p/d condition. The present paper provides the experimental results of the tube vibration response, hydrodynamic mass, damping ratio, and fluid-elastic instability. As the two-phase gap velocity increased, the fluidic-elastic instability occurred in the lift direction and a strongly coupled tube motion was found. The damping ratio was very dependent on the void fraction, as in the previous works. For a low void fraction flow, the fluid-elastic instability could be predicted by using Connors’ equation. However, the fluid-elastic instability in a high void fraction flow was quite different. The transition between the two fluid-elastic instability regions almost coincided with the flow regime transition criteria from a continuous bubbly flow to an intermittent flow.Copyright © 2006 by ASME