Computational fluid dynamics of advanced nuclear reactor cores and its contribution to safety analysis

• Main Author: Rowinski, Marcin Karol
• Other Authors: Zhao Jiyun
• Format: Theses and Dissertations
• Language: English
• Published: 2017
• Subjects: DRNTU::Engineering::Nuclear engineering; DRNTU::Engineering::Mechanical engineering::Fluid mechanics
• Online Access: http://hdl.handle.net/10356/72885
• Institution: Nanyang Technological University

Analytical methods and Computational Fluid Dynamics (CFD) examined the safety and performance of advanced nuclear reactors. Thermal and pressure drop evaluations of an innovative annular fuel in the European lead-cooled System ELSY fast reactor were conducted. Annular fuel has superior cooling capacity, and for a square fuel lattice, the pressure drop can be lower than the reference value for standard solid fuel. Consequently, when annular fuel rods replace solid fuel rods of the same dimensions the power rating is upgraded. While the hexagonal lattice cools more effectively, the pressure drop is higher than the standard fuel assembly and incompatible with power augmentation. Numerical investigations of supercritical flow of both carbon dioxide and water in a vertical tube under non-uniform heat flux applied at the wall demonstrated that current correlations for the Heat Transfer Coefficient (HTC) at low enthalpy values are accurate to ±15%. At high enthalpy, only the general trend is replicated due to large property value differences near the wall and in the bulk flow. In these circumstances, the thermal conductivity, specific heat, density and viscosity are 5-8 times lower after the transition from sub- to super-critical conditions. Moreover, the efficiency of different HTC correlation models is coolant dependant, e.g. the Swenson formula proves superior for carbon dioxide, whereas the Ornatsky model provides better agreement with water, where a lower maximum wall temperature was obtained when a non-uniform heat flux was applied, as compared to the reference case of uniform heat flux. For supercritical water flow inside a 2 2 fuel rod bundle, non-uniform heat flux increases wall temperature beyond the reference uniform flux. Also, wall temperature peaks occur only at the gaps between the fuel rods.