Measurement of fission gas release from irradiated UMo dispersion fuel samples

Journal Article · Thu Sep 01 00:00:00 EDT 2016 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2016.05.039· OSTI ID:1327093

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

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.

• Research Organization: Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
• Sponsoring Organization: USDOE
• DOE Contract Number: AC05-76RL01830
• OSTI ID: 1327093
• Report Number(s): PNNL-SA-114575; DN3001010
• Journal Information: Journal of Nuclear Materials, Vol. 478; ISSN 0022-3115
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

uranium-molybdenum
dispersion
fission gas release
nuclear fuel