• Nuclear
• Thermal-Hydraulic
• Subchannel CODE

CTF is the version of the thermal-hydraulic sub-channel code COBRA-TF (Coolant-Boiling in Rod Arrays-Two Fluids), designed for Light Water Reactor (LWR) analysis. It uses a two-fluid, three-field representation of two-phase flow, which makes the code capable of modeling two-phase flow in Boiling Water Reactors (BWR) during nominal operating conditions.
CTF has been subjected to a strict verification procedure, by addressing the mathematical accuracy of the numerical solutions on multiple stages. The code was then validated using numerous experimental databases, including the U.S. Nuclear Regulatory Commission (NRC) / Nuclear Energy Agency of the Organization for Economic Co-operation and Development (NEA-OECD) Boiling Water Reactor Full Bundle Tests (BFBT) Benchmark.
This degree thesis work is devoted to a further assessment of the CTF Code in comparison with experimental data from the BTBT, aiming for best- estimate evaluations of the nuclear reactor safety margins.
The BFBT benchmark contains a large amount of test cases representative of BWRs steady-state and off-nominal operating conditions, which makes it one of the most widely used benchmark for validating BWR modeling tools. One of the main experimental tests involve void distribution tests. Specific experimental cases were chosen and simulated using CTF. Statistical studies were carried out on the void distribution cases to evaluate the code modeling uncertainties.