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Title: Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested in INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-richmore » precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.« less
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  1. Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Fuels and Materials Division
Publication Date:
OSTI Identifier:
Report Number(s):
Journal ID: ISSN 2196-2936; PII: 55
Grant/Contract Number:
Accepted Manuscript
Journal Name:
Metallurgical and Materials Transactions. E, Materials for Energy Systems
Additional Journal Information:
Journal Volume: 2; Journal Issue: 3; Journal ID: ISSN 2196-2936
ASM International
Research Org:
Idaho National Laboratory, Idaho Falls, ID (United States)
Sponsoring Org:
Country of Publication:
United States
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS electron microscopy; nuclear fuel; research reactor; uranium molybdenum alloy; microstructure