In-pile transient testing of fuel segments with DIN 1.4970 cladding

Transient testing of fast reactor type fuel pins with DIN 1.4970 cladding has been performed in the TRIGA-ACPR Reactor at INR, Pitesti. The fuel test segments fabricated at SCK CEN resemble the fuel pins of the MYRRHA core, which is a lead-bismuth cooled fast reactor under development. The goal of the tests was to determine the cladding deformation as a function of the energy deposited in the UO₂ fuel, and to identify a potential cladding failure threshold (expressed in energy deposition) as a result of Pellet Cladding Mechanical Interaction (PCMI). Based on the reactor physics assessment and the test fuel segment design, a dedicated new irradiation rig was constructed. Twenty UO₂ fueled test segments were manufactured at SCK CEN using recently fabricated DIN 1.4970 cladding tubes with two different degrees of cold working. The un-irradiated test segments have been designed such that they resemble, as much as possible, fuel pins at a high burnup level. This has been achieved by reducing the pellet-cladding gap size and a higher level of the cladding cold-working degree. A total of 8 irradiation tests were performed in the TRIGA-ACPR on groups of 2 or 3 pins, that were subjected simultaneously to a short power pulse. The deposited power in the fuel and the temperature of the cladding was recorded during the test and cladding profilometry was measured before and after.