Steady state local temperature fields with turbulent sodium flow in nominal and disturbed bundle geometries with spacer grids

The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulic computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). These are the essential results obtained: outside the spacer grids the azimuthal temperature variations of the side and corner rods are greater by approximately the factor 10 in the bundle geometry under consideration as compared to rods in the central bundle zone. The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. Continuous reduction of a subchannel by rod bowing results in subtantial rises of temperature which, however, are limited to the adjacent cladding tube zones