Development of an innovative spacer grid model utilizing computational fluid dynamics within a subchannel analysis tool

In the past few decades the need for improved nuclear reactor safety analyses has led to a rapid development of advanced methods for multidimensional thermal-hydraulic analyses. These methods have become progressively more complex in order to account for the many physical phenomena anticipated during steady state and transient Light Water Reactor (LWR) conditions. The advanced thermal-hydraulic subchannel code COBRA-TF (Thurgood, M. J. et al., 1983) is used worldwide for best-estimate evaluations of the nuclear reactor safety margins. In the framework of a joint research project between the Pennsylvania State University (PSU) and AREVA NP GmbH, the theoretical models and numerics of COBRA-TF have been improved. Under the name F-COBRA-TF, the code has been subjected to an extensive verification and validation program and has been applied to variety of LWR steady state and transient simulations. To enable F-COBRA-TF for industrial applications, including safety margins evaluations and design analyses, the code spacer grid models were revised and substantially improved. The state-of-the-art in the modeling of the spacer grid effects on the flow thermalhydraulic performance in rod bundles employs numerical experiments performed by computational fluid dynamics (CFD) calculations. Because of the involved computational cost, the CFD codes cannot be yet used for full bundle predictions, but their capabilities can be utilized for development of more advanced and sophisticated models for subchannel-level analyses. A subchannel code, equipped with improved physical models, can be then a powerful tool for LWR safety and design evaluations. The unique contributions of this PhD research are seen as development, implementation,