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Title: Measurement of fission gas release from irradiated UMo dispersion fuel samples

The uranium-molybdenum (U-Mo) alloy dispersed in an Al-Si matrix has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. In this paper, two irradiated samples containing 53.6 vol% U-7wt% Mo fuel particles dispersed in an Al-2wt% Si matrix were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Measurements revealed three distinct fission gas release events for the samples from 400 to 700 oC, as well as a number of minor fission gas releases below and above this temperature range. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature with exceptional agreement.
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Publication Date:
OSTI Identifier:
Report Number(s):
Journal ID: ISSN 0022-3115; DN3001010
DOE Contract Number:
Resource Type:
Journal Article
Resource Relation:
Journal Name: Journal of Nuclear Materials; Journal Volume: 478
Research Org:
Pacific Northwest National Laboratory (PNNL), Richland, WA (US)
Sponsoring Org:
Country of Publication:
United States
uranium-molybdenum; dispersion; fission gas release; nuclear fuel