An experimental investigation is performed in a turbulent flow in a seven wire-wrapped rod bundle, mounted in an open air facility. Static pressure distributions are measured on central and peripheral rods. By using a Preston tube, the wall shear stress profiles are experimentally obtained along the perimeter of the rods. The geometric parameters of the test section are P/D=1.20 and H/D=15. The measuring section is located at L/D=40 from the air inlet. It is observed that the dimensionless static pressure and wall shear stress profiles are nearly independent of the Reynolds number and strongly dependent of the wire-spacer position, with abrupt variations of the parameters in the neighborhood of the wires.

Fuel elements of PWR type nuclear reactors consist of rod bundles, arranged in a square array, and held by spacer grids. The coolant flows, mainly, axially along the rods. Although such elements are laterally open, experiments are performed in closed type test sections, originating the appearance of subchannels with different geometries. In the present work, utilizing a test section of two bundles of 4x4 pins each, experiments were performed to determine the friction and the grid drag coefficients for the different subchannels and to observe the effect of the grids in the crossflow, in cases of inlet flow maldistribution.

In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies performed with various values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one.

The fuel element of LMFBR consists of a bundle of rods wrapped with an helical wire as spacer, surrounded by an hexagonal duct. In the present work, a semi-empirical model is developed to calculate bundle average and subchannel based friction factors and flow redistribution. The obtained results were compared to experimental data and they were considered satisfactory for wide range of geometrical parameters.