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The annular flow two-phase regime appears in nuclear power plants cooling by water. On accidents with lost of cooling this regime determine the heat extraction conditions from the core.  The experimental facility Gerador de Película Ondulatoria (GEPELON) was constructed at Universitat Politècnica de València to study this phenomenon.  In this work, we use computational fluid dynamics techniques to simulate an air-water regime, specifically a gravity-driven falling liquid film in the GEPELON experimental facility. We consider a multibody geometry where the porous sintered of the input device is simulated in a realistic way. The Volume of Fluid (VOF) model was used in the resolution of the constitutive equations. We propose three geometric approximations to estimate film thickness, and simulation results were compared with experimental values. The simulation results show a good agreement with experimental measurements. Besides, no significant differences were observed between geometrical approximations for film thickness calculations.

The use of advanced generation III+ and IV nuclear reactors, and their applications, has become important, seen as a means capable of contributing to the global transition to more sustainable, affordable and reliable energy systems. This technology, which could be integrated into future carbon-free electric power generation systems with high proportions of different renewable energy sources, includes Small Modular Reactors (SMR). There are about 100 different proposed projects of Generation III+ and IV, of which about 50 are SMR concepts, in various stages of development and of different types of technologies. Other important issues for achieving the long-term sustainability of nuclear energy are the proper use of its fuel sources and the improvement of nuclear waste management. Therefore, fuels based on a mixture of oxides have been used successfully in several countries. In addition, the incorporation of thorium-based fuel is a current challenge for the new designs of advanced reactors. The present paper focuses on the analysis of a small modular integral pressurized water reactor (iPWR) with Thorium-Uranium Oxide (Th-U MOX) mixtures. A thermohydraulic model is developed using the Ansys CFX program, which allows the calculation of the temperature distribution in the section where the highest power is produced within the SMR IPWR core (critical section). The temperature distributions in the fuel, clad and coolant were calculated with the objective of verifying that they were within the safety limits.

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The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this pers-pective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its   inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). We obtain the temperature profile distribution in the core for regimes where the coolant flow rate is smaller than recommended in a normal operation. In general, the temperature distributions calculated are consistent with phenomenological behavior. Even without considering the reactivity changes to reduce the reactor power or other safety mechanisms, the maximum temperatures do not exceed the recommended limits for TRISO fuel elements.