Ring compression tests on un-irradiated nuclear fuel rod cladding considering fuel pellet support

Ring compression tests on un-irradiated nuclear fuel rod cladding were conducted to analyze its mechanical behavior under pinching loads using appropriate boundary conditions, such as the consideration of fuel pellet support. The tested specimens included as-fabricated and artificially hydrogen-charged Zircaloy-4 cladding samples with a hydrogen content in a range of 285-470 ppm. Part of the samples were subjected to radial hydride treatment including a peak cladding hoop stress of 79 MPa and a peak cladding temperature of 400 °C to simulate SNF vacuum drying and to investigate treatment effects on the cladding response. Half of the tested rings were loaded with 20 mm long stainless steel pellets to analyze the impact of fuel pellet presence on the cladding loading capacity. The pellet-cladding gap width ranged from 60 to 180 μm. The other half of the rings was tested in empty state. The load-displacement curves obtained from ring compression tests conducted at room temperature on as-fabricated cladding exhibited a highly ductile material behavior. The presence of hydrogen in the cladding significantly embrittled the material, but unexpectedly, radial hydride treatment increased the cladding ductility. The ring compression tests conducted under pellet presence did not induce cladding cracking, even under extreme pinching loads. The results indicate that hydride-related material embrittlement likely does not cause nuclear fuel cladding failure when subjected to pinching loads, under the premise that a fuel pellet provides sufficient support to the cladding.