Accelerated fission rate irradiation design, pre-irradiation characterization, and adaptation of conventional PIE methods for U-10Mo and U-17Mo

Abstract

Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 1020 and 1.5 × 1021 fissions per cubic centimeter. Swelling measurement of the specimen sizes studied here required development and assessment of new methods for volume determination before and after irradiation. Laser profilometry and X-ray computation tomography (XCT) were used to provide preirradiation characterization of samples to determine the error and applicability of each to determine swelling following irradiation. These outcomes are contextualized through use of BISON simulations performed to assess the predicted expansion of U-Mo fuels subjected to the irradiation conditions of this work. Use of existing BISON fuel performance models predicted a maximum of 7% swelling under the irradiation conditions of this study. Pre-irradiation characterization revealed the as-cast U-Mo fuel samples were uniformly large-grained fully cubic U crystals with small U-C/N bearing precipitates and pores distributed throughout. Samples were found to contain a bulk porosity between .4 and 3% because of the casting process. Local porosity in areas far from large, interconnected pores was found by Slice-and-View to be under .2%. Nanometer-sized precipitates rich in C and N were identified in all samples, likely because of impurities during the fabrication process. Dendritic bands were also observed throughout the samples. These bands were characterized by variable Mo content that deviated from the overall Mo content by 2–3 wt %. No other microstructural features were correlated to these bands. Mechanical properties were found to be slightly strengthened compared to literature reports of bulk U-Mo fuels due to the nano-scale precipitates throughout the sample.

Nuclear fuel irradiation testbed for nuclear security applications

“…MiniFuel targets are shipped to the ORNL Irradiated Fuels Examination Laboratory (IFEL) for disassembly and PIE Raftery et al (2018); Harp et al (2021). The priority PIE is determined by the objectives of the experiment; this could be swelling, fission gas release, or evaluation of microstructural evolution under burnup in the case of more conventional fuel irradiations Doyle et al (2023), or other measurements as dictated by the goals of forensics, nonproliferation, or other nuclear security projects. Examples of the latter category of experiment are given in the following section.…”

“…Two specific examples of ongoing nuclear security and nonproliferation projects at ORNL are discussed herein: the intentional forensics project investigating the use of 'intentional taggants' for material identification purposes and the irradiated UO 2 material production for investigating the signatures of various isotopes that are produced in fuel as a function of enrichment and irradiation time. The testing of U-10Mo fuel for down-conversion of highly enriched reactors is also underway using this testbed but was omitted from this manuscript, see Doyle et al (2023) for more information on that work.…”