A whole-rod test was conducted at 229 °C to investigate the long-term stability of light water reactor spent fuel rods with artificially induced defects stored in inert and unlimited air atmospheres. Both boiling water reactor (BWR) and pressurized water reactor (PWR) rods were tested. After 2235 h, visual observations, diametral measurements, and radiographic smears were used to assess the degree of cladding deformation and particulate release. The same examinations plus metallography and x-ray analysis were conducted after 5962 h. Neither of the breached rods tested in inert atmosphere, nor the breached PWR rod tested in unlimited air, showed any measurable change from the pretest condition. The upper defect on the BWR rod tested in unlimited air had a 12.7-mm split after 2235 h and had 10% cladding deformation. The crack grew to 63.5 mm after 5962 h. X-ray analysis indicated that the UO2 had oxidized to U3O8 at this defect. The difference in the behavior of the upper and lower defects is attributed to the accessibility of the air to the fuel due to the positioning of the defect with relation to the pellet/pellet interface.