Tyler J. Gerczak

Advanced electron microscopy study of fission product distribution in the failed SiC layer of a neutron irradiated TRISO coated particle

Tristructural isotropic (TRISO) coated particle fuel has been designed for application in hightemperature gas-cooled reactors (HTGR). TRISO particles for the HTGR fuel development effort underway at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) consist of a two-phase uranium oxide-uranium carbide (UCO) fuel kernel, a carbon buffer layer, an inner pyrolytic carbon (IPyC) layer, a SiC layer, and an outer PyC (OPyC) layer [1]. The first in a series of irradiation experiments (AGR-1) clearly shows release of certain metallic fission products, e.g., Ag and Pd, through intact TRISO coatings, with Cs generally well retained [1]. No significant chemical interaction was observed between Pd and SiC for UCO TRISO coated particles, which retained Cs [2].

A Challenge to Multivariate Statistical Analysis: Spent Nuclear Fuel

Nuclear fission accounts for most of the non-polluting, non-fossil-fuel electrical power in the world. Higher burnup of fuel – that is, using a given fuel bundle for a longer time to produce more power – reduces the uranium resources needed, greatly enhances the economics of nuclear electricity, and reduces the amount of spent fuel for disposal. However, as the fuel burnup progresses, the fission process builds up large atomic fractions of fission products, consisting of most elements in the central region of the periodic table, in the fluorite UO2 matrix; and, a fuel/clad interaction (FCI) layer forms at the interface between the oxide fuel and the Zircaloy cladding. Providing a scientific basis for understating fuel behavior in the high burnup regime requires detailed characterization of high-burnup urania fuel. We have used X-ray spectrum imaging (SI) in SEM and STEM to analyze high-burnup (irradiated for 7 eighteen month long cycles to average burnup of 72 GWd/MTU) fuel from the H.B. Robinson pressurized water reactor. Multivariate statistical analysis (MVSA) is irreplaceable for understanding the extremely complex chemistry found.

Fabrication of Natural Uranium UO2 Disks (Phase II): Texas A&M Work for Others Summary Document

The steps to fabricate natural UO2 disks for an irradiation campaign led by Texas A&M University are outlined. The process was initiated with stoichiometry adjustment of parent, U3O8 powder. The next stage of sample preparation involved exploratory pellet pressing and sintering to achieve the desired natural UO2 pellet densities. Ideal densities were achieved through the use of a bimodal powder size blend. The steps involved with disk fabrication are also presented, describing the coring and thinning process executed to achieve final dimensionality.

PIE on safety-tested AGR-1 compact 4-2-2

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energys Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.

Multi-tier Analysis of SiC Breaches in Safety-Tested AGR-1 TRISO Fuel Particles

Tristructural isotropic (TRISO) coated particle fuel development is being supported by the US Department of Energy, Office of Nuclear Energy. The development plan includes a series of irradiations to qualify TRISO fuel. The first irradiation, AGR-1, included fuel fabricated at Oak Ridge National Laboratory (ORNL) and irradiated in the Advanced Test Reactor at the Idaho National Laboratory (INL). The TRISO fuel design consisted of a uranium carbide/uranium oxide kernel surrounded by concentric coating layers of carbonaceous buffer, inner pyrolitic carbon (IPyC), silicon carbide (SiC), and outer pyrolitic carbon (OPyC). Particles were then over-coated with carbonaceous matrix material and pressed into a cylindrical compact, with each compact containing greater than 4100 TRISO particles. A total of 72 compacts were included in AGR-1 and the irradiation was completed in November 2009 after ~620 effective full power days [1].