Accelerated High-Temperature Tests with Spent PWR and BWR Fuel Rods Under Dry Storage Conditions

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.