Conference Agenda

In water-cooled nuclear reactors, restoring core cooling after a Loss-of-Coolant Accident (LOCA) is critical, typically achieved through reflooding. During this process, the peak cladding temperature (PCT) arises between the dispersed flow film boiling and high-temperature vapor flow stages. Accurate PCT prediction is vital for reactor safety, spurring extensive research into droplet-wall heat transfer during high-temperature collisions. The boiling regimes—distinctive heat transfer mechanisms determined by wall temperature—necessitate precise identification for reliable modeling. However, previous studies using either hydrodynamic or thermal visualization to classify boiling regimes often yield inconsistent criteria due to their reliance on single techniques. This study addresses these limitations by simultaneously capturing hydrodynamic behavior and thermal characteristics, enabling improved boiling regime identification and comprehensive analysis of heat transfer mechanisms. Experiments used a circular substrate with two visual fields: a transparent section for observing droplet dynamics and an infrared-opaque section for thermal footprint detection. Substrate temperatures ranged from 150°C to 600°C, with droplets at saturation temperature released under gravity at a Weber number of 50. Side-view imaging measured residence time, spreading diameter, and rebound dynamics, while bottom-view imaging quantified the contact area. Infrared thermometry provided spatial heat flux distribution and overall heat transfer effectiveness. With increasing wall temperatures, distinct transitions between nucleate boiling, bubbly boiling, oscillating boiling, fingering boiling, and film boiling were identified. The combined visualizations provided detailed insights into effectiveness variations across boiling regimes, improving the understanding of droplet-wall heat transfer mechanisms. These findings support enhanced PCT modeling, advancing nuclear reactor safety analysis.

The present study addresses the critical need for accurate void fraction predictions in the engineering design and safety assessment of space-related two-phase systems. It investigates the drift-flux correlation under microgravity conditions, ranging from bubbly to annular flow regimes. This study compiles 458 experimental void fraction data points, revealing that distribution parameters vary with flow conditions and that drift velocities are minimal under microgravity conditions. Existing drift-flux correlations are found inadequate for capturing these variations and lack a simple model for drift velocity in microgravity two-phase flow. To address these issues, a new drift-flux correlation is proposed, considering flow condition effects on asymptotic distribution parameters and incorporating effective body acceleration to account for drift velocity decay in annular flow. The new correlation demonstrates strong predictive capabilities when evaluated against the collected experimental data, offering a significant advancement for space applications.

U-bend geometries are commonly used flow restrictions in nuclear reactor systems. Two-phase flows through U-bend are quite different from those in straight pipes. However, there has been no systematic study about inverted U-bend effects on two-phase flows. In the present study, a new experimental database is established using the existing Purdue University separate-effects test facility, featuring a 25.4 mm inner diameter pipe and a U-bend with curvature to diameter ratio R_c/D of 9. Detailed data including void fraction, gas velocity and bubble diameter are measured with miniaturized four-sensor conductivity probes with pressure loss obtained using pressure transducers. Using the obtained experimental data, mechanistic models have been developed to characterize the U-bend effects, which include models for pressure loss, U-bend dissipation length, variance of void fraction σ² and bubble velocity. It is found that the Lockhart-Martinelli’s two-phase flow frictional loss correlation can be used to predict the experimental two-phase pressure drop across U-bend with some modifications. The U-bend strength can be represented by the variance of the void fraction which dissipates exponentially in the U-bend dissipation region. The dissipation lengths of U-bend effects under different test conditions are determined by the dissipation rate β. The bubble velocity models are related to the development of σ². A modified Froude number Fr_m derived from the two-fluid model momentum equation is used as a fundamental parameter in developing the modeling correlations for σ², β, U-bend dissipation length and bubble velocity. All the modeled parameters can generally be predicted within an accuracy of ±10%.

In Boiling Water Reactors (BWRs), steam–water annular flow occurs near the fuel rods and plays a significant role in the nuclear reactor safety since the dryout of the liquid film may lead to the burn out of the fuel rods. However, the direct visualization and detailed liquid film measurement of high-temperature and high-pressure steam–water annular flow have been highly challenging due to the extreme operating conditions of BWRs (285°C and 7 MPa). This study addresses this limitation by developing a novel HFC134a–ethanol annular flow system at lower temperature and pressure (40°C and 0.7 MPa), effectively simulating the steam–water annular flow under BWR conditions. The experiments of HFC134a–ethanol upward annular flow were conducted in a 5 mm inner diameter tube using the constant electric current method and high-speed camera to obtain the liquid film thickness and flow behavior. The flow characteristics including base, average, and maximum film thickness and height of disturbance waves were obtained. Previous predictive models for these flow characteristics were tested with our measurement results. Through this simulating method, we report flow behaviors in detail achieving significant insights into liquid film behaviors in the actual BWRs.

U-bends are commonly used as flow restrictions in nuclear reactor systems, where two-phase flows exhibit significantly different behavior compared to straight pipes. Despite their importance, the effects of U-bends on two-phase flows remain underexplored. Interfacial area concentration (a_i), a fundamental parameter in two-fluid models, govern the interfacial transfer. The interfacial area transport equation (IATE) provides a superior approach to modeling a_i changes compared to conventional flow regime-dependent methods. In this study, IATE closure models have been developed using experimental database from the Purdue University separate-effects test facility, which features a 25.4 mm inner diameter pipe and a U-bend with curvature to diameter ratio R_c/D of 9. Experimental data suggests strong correlation between the variance of void fraction and covariance of Random Collision (COV_RC) in the U-bend and U-bend dissipation region. A modified Froude number Fr_m is used to model COV_RC. While constant values of covariance of Turbulent Impact (COV_TI) are used based on experimental results. Models for bubble velocity and pressure loss can be found in a separate U-bend geometric effects study. Void fractions are then determined using the continuity equation. Conventional drift-flux models are used in the straight pipe sections. Model coefficients of different bubble interaction terms are determined by evaluating each region (i.e., vertical upward, U-bend, U-bend dissipation, vertical downward) using experimental data individually. The one-group interfacial area transport along the whole test section is then evaluated using all the above closure models. The evaluation shows that the models predict a_i development accurately, with deviations generally within ±15%.

This work presents a mechanistic model for estimating the local heat transfer coefficient (HTC) in annular two-phase flow. The model is derived using fundamental principles and validated using data from two experimental facilities with different flow configurations and working flu-ids. Liquid-film thickness measurements were conducted using refrigerant at the University of Wisconsin-Madison, while HTC measurements were taken at an MIT facility using water as the working fluid. Non-invasive techniques are used at both laboratories to ensure the flow field is not disturbed. The physics-based modeling performed in this work ensures heat transfer performance in annular flow applications can be predicted with confidence.