Nuclear Technology / Volume 177 / Number 2 / February 2012 / Pages 257-272
Technical Paper / Materials for Nuclear Systems / dx.doi.org/10.13182/NT12-A13370
After irradiation and cooling in a pool, spent nuclear fuel assemblies are either transported for wet storage to a devoted site or loaded in casks for dry storage. During dry transportation or at the beginning of dry storage, the cladding is expected to be submitted to creep deformation under the hoop stress induced by the internal pressure of the fuel rod. The thermal creep is a potential mechanism that might lead to cladding failure. A new creep model was developed, based on a database of creep tests on as-received and irradiated cold-worked stress-relieved Zircaloy-4 cladding in a wide range of temperatures (310°C to 470°C) and hoop stress (80 to 260 MPa). Based on three laws - a flow law, a strain-hardening recovery law, and an annealing of irradiation hardening law - this model allows the simulation of not only the transient creep and the steady-state creep, but also the early creep acceleration observed on irradiated samples tested in severe conditions, which was not taken into account in the previous models.
The extrapolation of the creep model in the conditions of very long-term creep tests is reassuring, proving the robustness of the chosen formalism. The creep model has been assessed in progressively decreasing stress conditions, more representative of a transport.
Set up to predict the cladding creep behavior under variable temperature and stress conditions, this model can easily be implemented into codes in order to simulate the thermomechanical behavior of spent fuel rods in various scenarios of postirradiation phases.