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Film boiling is an important phenomenon in the evaluation of an emergency core cooling system following a hypothetical loss of coolant accident in a nuclear reactor. This study investigates the effects of liquid subcooling, surface oxidation, and surface materials on the minimum film-boiling temperature . Quenching experiments were performed using stainless steel and zirconium (Zr) test samples. The samples were heated to a temperature well above then plunged vertically in various degrees of liquid subcooling pools. A visualization study using a high-speed camera was conducted to capture the quenching behavior. Additionally, surface characterization analyses including X-ray diffraction, scanning electron microscopy, and energy dispersive X-ray spectroscopy were performed to quantify the surface conditions. Results indicate that liquid subcooling has a strong influence on . The visualization study shows a very thin vapor formation around the test sample for higher subcooling pools which explains the enhancement in the heat transfer. It is observed from the surface characterization analyses that the variations in the surface condition of the stainless steel and Zr causes the vapor bubbles to depart differently in the nucleate boiling regime. Furthermore, the effect of surface oxidation is clearly noticeable in the Zr test sample compared to the stainless steel test sample due to the oxidation kinematic of each substrate material. It is found that the substrate thermophysical properties have a significant impact on . Comparing the bare substrates shows that for the same degrees of liquid subcooling pool, the value of for the Zr sample is ∼30°C to 60°C higher compared to the stainless steel sample. Moreover, increasing the degrees of liquid subcooling contributes to a significant increase in that varies between ∼50°C and 70°C for both samples. [ABSTRACT FROM AUTHOR]

As part of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Rod Bundle Heat Transfer (RBHT) project, an experimental study was performed to investigate the entrained droplet sizes and velocities in a rod bundle under reflood conditions. Experimental results were obtained from the U.S. Nuclear Regulatory Commission/The Pennsylvania State University RBHT test facility using advanced dual laser measurement systems that allow for the simultaneous measurement of droplet behaviors at two axial locations during reflood transients. The RBHT facility is highly instrumented and contains a 7×7 electrically heated bundle with dimensions matching those in commercial pressurized water reactors. The combination of the measurement capabilities of the RBHT facility and the choice of appropriate experimental conditions allows for the measurement of unique droplet size and velocity distributions under different transient reflood conditions. [ABSTRACT FROM AUTHOR]

• An comprehensive code evaluation and validation is carried out using COBRA-TF. • The predictions are validated by reflood data from the NRC/PSU RBHT test facility. • COBRA-TF agrees generally well with the RBHT data. • Prediction errors exist for specific thermal–hydraulic quantities, requiring further model improvement. Accurate prediction of the reflood transients during LOCA has long been a challenging task. One reason for this is the substantial cost incurred to perform large-scale reflood tests, which are formidable in experimental expenditure and intensive in technology as well as program management. Another reason preventing a comprehensive understanding of reflood transients is the high-level complexity involved in the two-phase flow mass and heat transfer processes, which makes the measurement extremely difficult and inefficient. The situation has been significantly improved with the design and operation of the Nuclear Regulatory Commission (NRC)/Pennsylvania State University (PSU) Rod Bundle Heat Transfer (RBHT) test facility. A variety of advanced instrumentations were developed and used at this facility and very high-resolution data has been obtained, especially for the liquid droplet field. In the current study taking advantage of the unique NRC/PSU RBHT data, an extensive and comprehensive code evaluation and validation is carried out using the thermal–hydraulic sub-channel analysis code COBRA-TF. The system parametric effects investigated include: the system pressure, inlet liquid subcooling temperature, inlet flooding rate and rod bundle power. A variety of thermal–hydraulic quantities predicted by the numerical code are evaluated including: the quench front propagation, cladding, spacer grid and vapor temperature variations, two-phase pressure drop, liquid droplet velocity and coolant void fraction. In addition, the prediction errors are presented for each of the quantities in great detail in order to have a comprehensive understanding of the code performance. In general, COBRA-TF agrees relatively well with the experimental data in terms of the overall quench front propagation, only predicting slightly earlier quench. However, under low inlet subcooling and high flooding rate conditions, significant discrepancies are observed. The comparison with rod bundle thermal–hydraulic parameters indicates that COBRA-TF is able to predict the cladding temperature before quench well within a 15% range. The spacer grid and vapor temperature predictions are within 20% error. While the overall droplet velocity prediction is found to be within 30% error, COBRA-TF is able to capture the decreasing trend for droplet velocity during reflood. In addition, the prediction of the dispersed flow film boiling void fraction shows that the code always under-predicts the void fraction. While the most prediction discrepancy is found to be the two-phase flow pressure drop after quench, which involves more than 50% under-prediction for the bulk liquid boiling regime. The present study clarifies the effect of different system parameters on the various two-phase flow quantities during reflood. The results obtained provide answers for many existing code modeling issues and thus can be instructive and useful for future model development and code upgrading. [ABSTRACT FROM AUTHOR]