In the framework of the liquid-metal fast breeder reactor safety analysis program, out-of-pile sodium boiling experiments have been run at Kernforschungszentrum Karlsruhe in a 37-pin bundle simulating a fast reactor subassembly. Three representative runs are analyzed in detail in terms of experimental evaluation and numerical simulation. The latter is performed with the three-dimensional, two-phase flow computer code BACCHUS-3D/TP, which describes coolant behavior in bundle geometry. The comparison between computed and experimental results has helped in correlating data from different instruments, thus allowing deeper insight into the details of the boiling behavior. Experimental data also provided a valuable code verification. By modifying the drift-flux model, the code validity range has been enlarged.