Experimental studies of pressurized-water-reactor flow and power transients are performed on a 6-ft-long electrically heated 9-rod bundle in a square array. Flow transients are patterned after decay rates typical of reactor coolant pump coast-downs. Power transients are approximately 5% ramp increases. For the transient conditions tested, there is no premature occurrence of critical heat flux. Steady-state critical heat flux data for axial spacer grid separations of 10, 15, and 21 in. indicated there is no grid-spacing effect on critical heat flux by the Babcock & Wilcox Mark B2-style spacer grid at normal reactor flow rates.