The mechanical integrity of the nuclear fuel cladding is critical for the safety of Pressurized Water Reactor (PWR) fuel elements, in normal and off-normal operating conditions. Stresses in the cladding result from many phenomena, but are mostly driven by the mechanical interaction of the cladding with the ceramic fuel. Hence, a comprehensive description of the mechanical behavior of the fuel is crucial, and many studies have been devoted to its understanding and description. This viscoplastic behavior is characterized by mechanical tests (uniaxial compression tests, bending tests) on a virgin material for reasons now discussed. Aside from the intrinsic constraints associated with testing irradiated materials (which requires usage of dedicated hot cells), a severe difficulty  is to extract suitable samples from irradiated fuel pellets, which are generally fragmented and highly damaged during the irradiation.

The objective of the work is to prove the feasibility of a mechanical test on irradiated ceramics. To do so, we will address all the technical challenges and present our innovative solutions in order to implement the world first creep experiment on spent fuel in a hot cell.

Initially, the extraction of an intact sample from cracked irradiated fuel conditions the feasibility of such a creep test. To lift this barrier, we explore the latest advances in ultrasonic non-destructive testing to map the cracks inside ceramics. This method will be compared with otherwise mastered destructive methods in order to establish the most robust reproducible process to lead the way for a future modus operandi. Then, the ceramics will be machined into millimeter thick cylindrical samples. The tests will be carried out in a unique induction furnace to reach high temperatures under a well-controlled atmosphere and equipped with a load line that is instrumented to control temperature, displacement and load.

The development of our device extends over three « twin platforms », making it possible to qualify each step independently, starting with testing inert materials, and then non-irradiated nuclear ceramics to compare with a large available database. Finally, we will introduce a robust experimental loop in a hot cell to set up the first database on spent fuel.

To support our experimental efforts, we use a micrometric 3D scanner technology to scan our experimental samples and integrate them in a digital twin. Such experimental devices shall allow identifying viscoplastic behaviors of irradiated fuels suitable to enrich the fuel performance codes using to this day post-irradiation indirect reverse methods.