This workshop gives an overview of the advanced thermal-hydraulic sub-channel code CTF and its nuclear fuel rod solver CTFFuel. The code is used for steady state and transient design and safety analyses of current and advanced nuclear reactors. CTF is also the thermal-hydraulic component of the CASL-developed Virtual Environment for Reactor Applications (VERA) core simulator used for thermal feedback and prediction of thermal-hydraulic safety parameters. The international CTF users’ group has currently more than 60 member-organizations from different countries including industry, regulation, national labs, consulting companies, research institutes and academia. The theory and physics models of CTF/CTFFuel will be presented first, followed by an interactive real-time demonstration on how to build fuel assembly and core models, execute the code in a serial and a parallel mode, and post process the results using different available tools.

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