In-Cheol Chu

Multidimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus During LBLOCA Reflood Phase with a Direct Vessel Injection Mode

Abstract Multidimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor (PWR) vessel with a direct vessel injection mode is presented based on the experimental observation in the MIDAS (multidimensional investigation in downcomer annulus simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a large-break loss-of-coolant accident (LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled down to a 1400-MW(electric) PWR type of a nuclear reactor, focused on understanding multidimensional thermal-hydraulic phenomena in a downcomer annulus with various types of safety injection during the refill or reflood phase of an LBLOCA. The initial and the boundary conditions are scaled from the pretest analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer.

Bubble Lift-off Diameter and Nucleation Frequency in Vertical Subcooled Boiling Flow

A series of experiments were carried out to investigate phenomena from bubble nucleation to lift-off for a subcooled boiling flow in a vertical annulus channel. A high-speed digital video camera was used to capture the bubble dynamics. The bubble lift-off diameter and bubble nucleation frequency were evaluated in terms of heat flux, mass flux, and degree of subcooling. The fundamental features of the lift-off diameter and nucleation frequency (i.e., the variations across nucleation sites and the dependence on the flow and heat flux conditions) were addressed based on the present observation. A database for the bubble lift-off diameter was built by gathering and summarizing the data of Prodanovic et al., Situ et al., and the present experiments. We evaluated the predictive capabilities of Unals model, Situ et al.s model, and Prodanovic et al.s correlation against the database. We obtained the best prediction results by modifying the wall superheat correlation in Unals model. In addition, we suggested a new correlation for a combined parameter of the bubble nucleation frequency and bubble lift-off diameter.

Fluid-Elastic Instability of Rotated Square Array U-Tubes in Air-Water Flow

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.

Integral Effect Tests on Transient Thermal-Hydraulic Behavior During a Steam Generator Tube Rupture Accident in the APR1400

Abstract A postulated steam generator tube rupture (SGTR) event of the APR1400 (Advanced Power Reactor 1400 MWe) was experimentally investigated with the thermal-hydraulic integral effect test facility ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation). The SGTR accident is one of the design-basis accidents having a significant impact on safety from the viewpoint of radiological release. To simulate a SGTR accident of the APR1400, the SGTR-HL-04 and the SGTR-HL-05 tests were performed by simulating double-ended ruptures of a single U-tube and five U-tubes at the hot side of the ATLAS steam generator. Following the reactor trip induced by a high steam generator level signal, the primary-system pressure decreased and the secondary-system pressure increased until the main steam safety valves were opened to reduce the secondary-system pressure. A mild change of the water level in the core was observed, which was attributed to the small break sizes of the present tests compared with conventional loss-of-coolant-accident tests. No excursion in the cladding temperature was observed in either test. The break area affected the timing of the major events observed in the tests. Lessened heat transfer to the secondary side caused by earlier actuation of the safety injection pumps had more influence on the secondary pressure of the affected steam generator than the break flow. The break flow was discharged as single-phase water in both tests.

Bubble Lift-Off Diameter in Forced Convective Boiling Flow

A series of experiments were carried out to investigate the bubble nucleation to lift-off phenomena for a subcooled boiling flow in a vertical annulus channel. A high speed digital video camera was used to capture the dynamics of the bubble nucleation to lift-off process. A total of 148 recordings were made, and the bubble lift-off diameter and the bubble nucleation frequency were evaluated for 118 recordings up to now. The basic features of the lift-off diameter and nucleation frequency were addressed based on the present observation. A database for the bubble lift-off diameter was built by gathering and summarizing the data of Prodanovic et al., Situ et al., and the present work. The prediction capability of Unal’s model, Situ et al.’s model, and Prodanovic et al.’s correlation was evaluated against the database. The best prediction results were obtained by modifying the wall superheat correlation in Unal’s model.Copyright

Flow-Induced Vibration Responses of U-Tube Bundle in Air-Water Flow

In the present study, a series of experiments have been performed to investigate a fluid-elastic instability of a nuclear steam generator U-tube bundle in an air-water two-phase flow condition. A total of 39 U-tubes are arranged in a rotated square array with a pitch-to-diameter ratio of 1.633. The diameter and other geometrical parameters of U-bend region are the same to those of an actual steam generator, but the vertical length of U-tubes are reduced to 2-span in contrast to 9-span of an actual steam generator. The following parameters were experimentally measured to evaluate a fluid-elastic instability of U-tube bundles in a two-phase flow: a general tube vibration response, a critical gap velocity, a damping ratio and a hydrodynamic mass. Based on the experimental measurements, the instability factor, K, of Connors’ relation was preliminary assessed with some assumptions on the velocity and density profiles of the two-phase flow.Copyright

Fluid-Elastic Instability in Tube Bundles and Effect of Flow Regime Transition

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