Release of fission tritium through Zircaloy-4 fuel cladding tubes

Abstract In order to analyse a potential fission tritium migration from the fuel to the coolant trough the cladding, two experiments of effusion and permeation types, have been performed on 3 H release from Zry4 claddings. During the tests at 350 °C, the 3 H released activities were measured at regular intervals. In both cases, very fast release rates have been obtained in the first few days, followed by more steady release rates. A correlation has been obtained between the 3 H releases measured and the oxide formation kinetics after the initial burst. A mechanism of 3 H transport is proposed based on the behaviour of the precipitates during the oxidation of Zry4. Applied to the conditions of PWR fuels, the measurements performed and the mechanisms considered lead to an insignificant contribution of fission tritium permeation to the total inventory of the tritium in the primary coolant.