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

Journal Article · Mon Jun 01 00:00:00 EDT 2015 · Journal of Nuclear Materials, 461:61-71

DOI:https://doi.org/10.1016/j.jnucmat.2015.02.020· OSTI ID:1184898

Burkes, Douglas; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine; Pool, Karl N.

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.

• Research Organization: Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
• Sponsoring Organization: USDOE
• DOE Contract Number: AC05-76RL01830
• OSTI ID: 1184898
• Report Number(s): PNNL-SA-101195; NN9100000
• Journal Information: Journal of Nuclear Materials, 461:61-71, Journal Name: Journal of Nuclear Materials, 461:61-71
• Country of Publication: United States
• Language: English

Related Subjects

Uranium-Molybdenum
Fission Gas
Thermal Property
Nuclear Fuel