Reactor Core Thermal-Hydraulics

Catalog description

In-depth analysis of the reactor core thermal hydraulics; computational methods and practical applications. NUC E 502 Reactors Core Thermal-Hydraulics (3) This course provides students with a background in reactor core thermal hydraulics and enhances their understanding of the important phenomena in a nuclear reactor core, which can determine reactor safety performance. Students will obtain an overall view of reactor safety from the reactor thermal hydraulics perspective. This course examines the outcomes of research projects and international scientific activities in this area. Objectives are met by introducing course modules that utilize state-of-the-art computer codes to solve well established international thermal-hydraulics benchmark problems to demonstrate reactor performance during operational transients. One of the principal goals of the course is to provide students with a computationally intensive curriculum that is consistent with their capabilities and their expectations for a modern reactor thermal hydraulics course. This course discusses detailed thermal-hydraulic analysis of reactor systems with an emphasis on the application of conservation equations for single- and two-phase flow in detailed modeling of reactor cores using three-dimensional subchannel analysis methods and examines the reactor's core thermal-hydraulic design for core limit analysis. The governing sets of equations that form the basis for the three-dimensional thermal-hydraulic methods commonly used in the nuclear industry will be derived and discussed in addition to specific models that are used for closure. Hot assembly analysis will be performed, as well as core wide analysis, to determine the hot assembly and resulting hot subchannels in the core. Students will use state -of-the-art three-dimensional computer codes to model fuel assemblies and the reactor core to determine the most limiting fuel pin and hottest subchannel. Background on heat and mass transfer and fluid dynamics is the prerequisite to this course, which provides a basis for understanding reactor core thermal-hydraulic analysis.