COBRA-TF (Coolant Boiling in Rod Arrays – Two Fluid) or CTF is a transient subchannel code, selected to be the reactor core thermal hydraulic (T/H) simulation tool in the multi-physics code development project of the Consortium for Advanced Simulation of Light Water Reactor (CASL) sponsored by the US Department of Energy. CTF is currently being evaluated and further improved by CASL as part of its multi-physics software package to help the nuclear industry address operational and safety challenge problems, such as Departure from Nucleate Boiling (DNB) and Reactivity Initiated Accidents (RIA). In this paper, CTF’s capability for transient fuel thermal analysis, including DNB prediction is evaluated by modeling and simulating power burst experiments with high burnup PWR fuel rods, conducted at the Nuclear Safety Research Reactor (NSRR) in Japan. The experiments were a series of tests performed using pulse irradiation capability of the reactor to evaluate fuel rod failure with respect to fuel enthalpy, coolant conditions, and fuel design during RIAs such as control rod ejection. Specific to this study, the experiments using the Takahama-3 reactor fuel segments have been modeled and simulated to evaluate CTF’s prediction capability for DNB onset, fuel rod thermal response, and heat transfer from single-phase to post-CHF during fast RIA transients.

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