Burst and Tensile Testing of Hydrided Fuel Cladding

The U.S. Department of Energy Office of Nuclear Energy, Office of Fuel Cycle Technology has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste. Within the UFDC, the storage and transportation task has been created to address issues of extended storage and transportation. One gap in knowledge identified as high priority by UFDC includes hydrogen effects in zirconium alloycladding and reorientation of clad hydrides. During reactor operations, hydrogen, generated from the oxidation of the surface of zirconium alloy materials in the coolant of light water reactors, can be absorbed into the cladding and form zirconium hydrides. At high temperatures and hoop stresses, the normally circumferential-oriented hydrides in High Burnup UNF (HBUNF) may reorient in the radial direction and affect the mechanical properties of the clad. Understanding these changes in mechanical properties will aid in analyzing how UNF will perform during storage and transportation. Separate effects testing to measure the mechanical properties of un-hydrided, pre-hydrided, and irradiated HB-UNF is currently underway at Pacific Northwest National Laboratory (PNNL). In order to measure these mechanical properties, staff at PNNL are employing tensile, compression, and burst testing. Each of these test stands are capable of testing the clad at temperatures up to 400oC. The development of these test stands and future testing of pre-hydrided and HB-UNF cladding are described herein.