The long-term management of spent nuclear fuel could, in France, involve transportation of spent fuel assemblies from one spent fuel pool to another, possibly ending with the assemblies located in dry storage. This situation would lead to some thermomechanical cycling, with possible effects on the fuel cladding's fracture resistance. An experimental study was performed at IRSN on nonirradiated Zircaloy-4 cladding tubes that were prehydrogenated in order to be representative of the hydrogen content of irradiated cladding. The specimens were “C”-shaped ring segments subjected to a compression load, generating a wide range of stress levels. Radial hydride treatment (RHT) consists of heating the samples to a maximum temperature of 350 or 400°C, and then cooling them at 0.4°C/min under constant compression load to room temperature. The use of a constant compression load guaranties a constant stress at any given location of the sample. The influence of a given stress level on radial hydride precipitation can consequently be evaluated at various locations in the ring to determine two possible stress thresholds: the stress below which no radial hydride precipitation is observed and the stress level above which only radial hydride precipitation is observed. Another interesting outcome is the stress level required at room temperature to nucleate a crack along a radial hydride after RHT. Load-displacement records helped to distinguish between brittle and ductile crack nucleation, which was then followed by crack instability and ductile tearing. In the present study, a large set of tests addressing the influence of hydrogen content with: 2, 5, 10, and up to 30 RHT cycles was performed at two maximum temperatures, 350 and 400°C, and these were compared to the results of formerly obtained single RHT cycle tests. The obtained results provide insights for a better understanding of the expected influence of thermomechanical cycling on irradiated cladding.