Recent Advances in Computational Convective Heat Transfer Study in a Sub-Channel for Nuclear Power Reactor and Future Directions

Abstract

A nuclear power reactor's primary use is to generate thermal energy, which in turn produces electricity. The primary heat source is a nuclear fission event occurring inside the fuel rod. The convection heat is transmitted through the coolant by the heat energy generated at the fuel rod wall boundary. Better heat transfer is produced in the flow area by turbulence and irregularity. As a result, turbulent flow heat transfer may present a significant challenge when predicting and assessing the thermal performance of nuclear power reactors. Computational techniques in convective heat transfer have become indispensable for solving challenging issues in the fields of science and engineering thanks to the development of current sophisticated numerical methods and high-performance computer hardware. The development of novel computational techniques and models for complicated transport and multi-physical phenomena is constantly in demand throughout applicable disciplines. This chapter's objective is to provide some recent developments in computational techniques for convective heat transfer, taking into account research interests in the community of mass and heat transfer, and to showcase relevant applications in nuclear power plant engineering domains including future directions. This study describes the most recent advancements in nuclear reactor convective heat transfer research utilizing the computational fluid dynamics (CFD) method, particularly at Ansys Fluent. This work examines the convective heat transfer and fluid dynamics fluid dynamics for turbulent flows across three rod bundle sub-channels that are typical of those employed in the PWR-based VVER type reactor. In this paper, CFD analysis is carried out using the software tool Ansys Fluent. Temperature distribution profile, velocity profile, pressure drop, and turbulence properties were investigated in this study. Boundary conditions i.e. temperature, velocity, pressure, heat flux, and heat generation rate were applied in the sub-channel domain. The main obstacles and bright spots for the CFD methods in nuclear reactor engineering are discussed, which helps to further its further uses. We intend to research a full-length fuel bundle model for VVER-1200 in the future to gather specific fluid characteristic data and use the findings to analyze safety and operate nuclear power facilities in Bangladesh. This paper presents a thorough analysis of the sub-channel thermal hydraulic codes used in nuclear reactor core analysis. This review discusses several facets of previous experimental, analytical, and computational work on rod bundles and identifies potential future directions based on those earlier studies.