Accelerated high-temperature tests on 25 intact pressurized water and boiling water reactor rods were conducted for more than 16 months at 400, 430, and 450°C in a helium gas atmosphere. The pretest characterized rods were examined by nondestructive methods after each of the three test cycles. No cladding breaches occurred and the creep deformation remained below 1%, which was in good agreement with model calculations. The test atmospheres were analyzed for 85Kr and tritium. The 85Kr concentrations were negligible and the tritium release agreed with the theoretical predictions. It can be concluded that for Zircaloy-clad fuel, cladding temperatures up to 450°C are acceptable for dry storage in inert cover gases.