Search

• A new two-phase flow mass quality correlation was developed. • One-dimensional conservation equations were formulated for liquid and vapor phases. • Scaling analysis was performed to obtain an expression for the vapor drift velocity. • The new correlation predicts the mass quality in DFFB regime with improved accuracy. A new semi-empirical two-phase flow mass quality correlation was developed for the post-dryout dispersed flow film boiling (DFFB) regime in a 7 × 7 rod bundle geometry. Starting from the fundamental physics describing the two-phase flow in the dispersed droplet flow regime, one-dimensional conservation governing equations were formulated for the liquid and vapor phases by considering a detailed characterization of the liquid droplet phase, based on which the liquid phase velocity was related to the key flow parameters. A scaling analysis was then performed within a sub-channel of the rod bundle to obtain an expression for the vapor drift velocity, from which a suitable relationship for the two-phase flow quality was derived using the drift flux model. It was found that the vapor drift velocity in the DFFB regime is affected by many variables including: the droplet size, velocities of the liquid and vapor phases, void fraction, vapor superheat and fluid properties, all of which are important in characterizing the dispersed droplet flow. Based on the results of the RBHT reflood tests, which simulate the reflood transients of a LWR during an accident scenario, a new mass quality correlation was developed that is applicable to the conditions with system pressure ranging from 138 to 414 kPa, rod bundle peak power between 0.98 and 1.97 kW/m, inlet liquid subcooling temperature from 11 to 83 K, and inlet flooding rate ranging from 0.0191 to 0.0508 m/s. Comparisons with experimental data indicated that the new correlation is able to predict the actual mass quality in the post-dryout two-phase flow regime with significantly improved accuracy not only for the rod bundle geometry but also for the tube geometry. [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]

• A new spacer grid pressure drop correlation is developed for the DFFB regime. • The model is validated by reflood data obtained from the NRC/PSU RBHT test facility. • The correlation improves the spacer grid pressure drop predication within a 25% error span. Flow blockages such as the spacer grids play an important role in a two-phase flow system as they modify the flow and heat transfer behavior. Their existence, on the one hand, increase the turbulent mixing and heat transfer capability of fluid directly downstream; while on the other hand, create additional pressure loss within the system thus requiring a detailed and comprehensive understanding. In the present work, a new two-phase flow pressure drop correlation induced by spacer grids in the dispersed flow film boiling (DFFB) regime is developed based on the unique reflood experimental data obtained from the Nuclear Regulatory Commission/Pennsylvania State University rod bundle heat transfer test facility. The test conditions cover wide ranges of system parameters for two-phase flow system which can be used for model development. It is found that the most dominating system parameter is the mixture mass flow rate across the spacer grid location. As a result, the spacer grid loss coefficient is correlated with the mixture Reynolds number and a blockage geometry correction factor. The current correlation is found to significantly improve the prediction of the spacer grid pressure drop in the DFFB regime generally within a 25% error span, while the other exiting correlations failed to capture the two-phase flow hydraulic behavior involved during the reflood transients. The current correlation developed can be readily implemented into existing thermal-hydraulic analysis codes to further improve their prediction capabilities in the DFFB regime with the presence of flow blockages. [ABSTRACT FROM AUTHOR]

• A new spacer grid induced droplet breakup correlation was developed. • Conservation of mass, energy, and important flow quantities are considered. • The NRC/PSU RBHT reflood data is used for model development. • New correlation predicts the droplet size after breakup within a 17% error span. • New correlation predicts the validation data set with RMSE of 0.1060 and standard deviation of 0.0990. The liquid droplet breakup process at the leading edge of a dry spacer grid during reflood transients is studied theoretically based on the experimental observations in the present work. Various rod bundle quenching experiments were carried out at the Rod Bundle Heat Transfer (RBHT) test facility to simulate the reflood transients of a Loss of Coolant Accident (LOCA) for Light Water Reactors. In the experiment, variation of the droplet size distributions across a spacer grid are measured and analyzed. The size of droplet generated downstream of a spacer grid is found to be affected by the condition of the spacer grid as well as the quench front propagation. In addition, the substantially reduced droplet size at downstream of the spacer grid indicates that droplet breakup takes place. Based on available experimental data obtained in the Rod Bundle Heat Transfer tests and in literature, a theoretical model incorporating two droplet groups (large and small droplet groups) is developed for the dry spacer grid droplet breakup by considering the conservation of liquid mass, energy transport, and including various important flow and heat transfer parameters including the Nusselt number, Reynolds number, Jakob number, Weber number, dimensionless radiation number, distance from the initial breakup point, incoming droplet size, and spacer grid blockage ratio. Moreover, a liquid droplet mass loss coefficient was proposed for small droplet group to specially account for the liquid mass evaporation due to interfacial heat transfer and radiation. The model so developed is found to be able to predict the RBHT data within 17% error with RMSE of 0.1060 and standard deviation of 0.0990 for the validation data set, showing significantly increased accuracy. Comparison made with other experimental results indicates that the current model developed correctly captures the mass and energy transport process during breakup. In addition, sensitivity analyses are performed to investigate the overall performance and relative importance of various parameters that are related to the breakup process in the current study. The current model developed can be incorporated into nuclear reactor thermal-hydraulic and safety analysis codes to enhance their capabilities of predicting the droplet behaviors and the fuel rod peak cladding temperature in the dispersed flow film boiling regime during reflood transients of postulated accident scenarios such as LOCA for light water reactors. [ABSTRACT FROM AUTHOR]