COBRA-TF (Coolant Boiling in Rod Arrays–Two Fluid), or CTF, is a transient subchannel code selected to be the reactor core thermal-hydraulic simulation tool in the multiphysics code-development project of the Consortium for Advanced Simulation of Light Water Reactor (CASL) sponsored by the US Department of Energy (DOE). In this paper, CTF’s capability for departure from nucleate boiling (DNB) prediction is evaluated by modeling and simulating power-burst experiments with high-burnup pressurized water reactor (PWR) fuel rods, conducted at the Nuclear Safety Research Reactor (NSRR) in Japan. Experiments using reactor fuel segments have been modeled and simulated to evaluate CTF’s prediction capability for onset of DNB and heat transfer from single-phase to post-critical heat flux (CHF) during fast reactivity-initiated accident (RIA) transients. The calculations demonstrated that CTF is able to simulate a fast transient with a large power pulse. CTF predicted DNB occurrence in all of the cases, after the power pulse, consistent with experimental observations. In the experiments, all cases experienced DNB, as predicted by the correlations in CTF; however, fuel failure occurred in only two of the cases: one at the peak power and the other after the peak power. The remaining cases survived with enthalpies significantly higher than those that failed while experiencing DNB occurrences.