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Title: Measurement of Fission Gas Release from Irradiated U-Mo Monolithic Fuel Samples

The uranium-molybdenum (U-Mo) alloy in a monolithic form 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. An apparatus capable of annealing post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1050 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in literature.
Authors:
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Publication Date:
OSTI Identifier:
1184898
Report Number(s):
PNNL-SA-101195
NN9100000
DOE Contract Number:
AC05-76RL01830
Resource Type:
Journal Article
Resource Relation:
Journal Name: Journal of Nuclear Materials, 461:61-71
Research Org:
Pacific Northwest National Laboratory (PNNL), Richland, WA (US)
Sponsoring Org:
USDOE
Country of Publication:
United States
Language:
English
Subject:
Uranium-Molybdenum; Fission Gas; Thermal Property; Nuclear Fuel