The behaviour of defected zircaloy-clad UO2 fuel elements with graphite coatings between fuel and sheath irradiated at linear power of 48 kW/m in pressurized water

Three purposely defected Zircaloy-clad UO2 fuel elements were irradiated separately in a pressurized water loop at linear powers of 48 kW/m for times up to 850 hours. The steady state and transient release of gaseous fission products to the loop coolant were measured with an on-line γ-ray spectrometer. During irradiation each element lost UO2 from the pellet under the defect hole in the sheath; this loss of UO2 was attributed to grain boundary oxidation. The fractional steady state release of fission products ranged from 1.4 x 10-1 for the longer-lived xenons to 3.5 x 10-5 for the shorter-lived Xe-138, indicating a delay between birth and release, either in the UO2 or in the fuel-to-sheath gap. Release of fission products on reactor power transients were up to four times higher than the steady state release. Release values calculated by two different models were in reasonable agreement with experimental results. (auth)