In this work, the sub-channel thermal-hydraulic code CTF is applied to the hottest fuel assembly of a VVER-1000 core, aiming to investigate the code sensitivity to uncertainties of the initial and boundary conditions. The core thermal-hydraulic solver CTF is a modernized version of the COBRA-TF sub-channel code, which is being maintained and developed by the Reactor Dynamics and Fuel Modeling Group (RDFMG) at North Carolina State University (NCSU) in cooperation with Oak Ridge National Laboratory (ORNL).

In this study, first, a full core model of a VVER-1000 reactor with its initial loading pattern is created for the Monte Carlo neutronics code MCNP6 under normal operating conditions using ENDF/B VII.1 / NJOY99. The assembly power factors and the pin-powers of the hottest fuel assembly, obtained by MCNP6, are used as power boundary conditions in CTF. The hottest assembly is simulated to calculate the fuel, cladding, and coolant temperatures at normal operating conditions.

Uncertainty analyses are performed using Dakota 6.5 and it is observed that CTF predictions of fuel, cladding, and coolant temperatures are most sensitive to uncertainties in core average power and inlet coolant temperature.