Isothermal Shakedown Tests for the 5×5 Pressurized Water Reactor Rod Bundle in the Automated Critical Heat Flux Test Facility

Abstract

Pressurized water reactors employ rod bundles to generate thermal power via nuclear fission – which is extracted using water as the process fluid. Along the length of the rod arrangement, spacer grids are installed to increase turbulence and heat transfer rate. It is critical to estimate the temperature variations, pressure drop and frictional losses to obtain a better understanding of the efficiency and safe operation of a particular design used. The TAMU CHF Facility is a complex experimental facility designed with the intent to study such operations. Development of such a facility is imperative to study thermal hydraulic parameters for specific fuel rod and spacer grid designs, and verify the results with computer modeled simulations. The facility was upgraded to facilitate automation, remote operation, acquisition of high-fidelity experimental data and analysis. Subsequently, temperature changes, pressure and flow related parameters were estimated at various Reynolds numbers for specified increments in loop pressure, using water as the process fluid.