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. © 2017 Elsevier Inc. All rights reserved.