Sc23667-htwr.part4.rar

Analysis of fuel rod material behavior at high temperature, referencing material-specific thermal conductivity plots.

Inlet temperature, pressure, and mass flow rates derived from experimental data. 3. Results and Discussion sc23667-HTWR.part4.rar

The study demonstrates that the SC23667 design meets safety standards for core thermal limits during transients. The developed numerical codes show high accuracy in predicting thermal-hydraulic phenomena within the reactor core. Analysis of fuel rod material behavior at high

Application of Navier-Stokes equations with turbulence modeling in modern thermal-hydraulic codes. sc23667-HTWR.part4.rar