Abstract
The DOE Used Fuel Disposition Campaign
(UFDC) has tasked ORNL to investigate the behavior
of light-water-reactor fuel cladding material
performance related to extended storage and
transportation of used fuel. Fast neutron irradiation
of pre-hydrided zirconium-alloy cladding in the High
Flux Isotope Reactor (HFIR) at elevated
temperatures has been used to simulate the effects of
high burnup on used fuel cladding for use in
understanding the materials properties relevant to
very long-term storage (VLTS) and subsequent
transportation. The irradiated pre-hydrided metallic
materials will generate baseline data to benchmark
hot-cell testing of high-burnup used fuel cladding;
and, more importantly, samples free of alpha
contamination can be provided to the researchers
who do not have hot cell facilities to handle highly
contaminated high-burnup used fuel cladding to
support their research projects for the UFDC.
In order to accomplish this research, ORNL has
produced unirradiated zirconium-based cladding
tubes with a certain hydrogen concentration. Two
capsules (HYCD-1 and HYCD-2) containing
hydrided zirconium-based samples, 9.50 mm (0.374
in) in diameter, were inserted in HFIR for neutron
irradiation. HYCD-1 was removed after Cycle 440B
and HYCD-2 after Cycle 442. This paper will
describe the general HYCD experiment
configuration, achieved temperatures, and
temperature gradients within the cladding, and
current results of the PIE of the irradiated hydrided
cladding samples.
