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International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver Downtown
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Argonne’s METL gears up to test more sodium fast reactor components
Argonne National Laboratory has successfully swapped out an aging cold trap in the sodium test loop called METL (Mechanisms Engineering Test Loop), the Department of Energy announced April 23. The upgrade is the first of its kind in the United States in more than 30 years, according to the DOE, and will help test components and operations for the sodium-cooled fast reactors being developed now.
Kazuki Hida, Sadao Kusuno, Takeshi Seino
Nuclear Technology | Volume 75 | Number 2 | November 1986 | Pages 148-159
Technical Paper | Fuel Cycle | doi.org/10.13182/NT86-A33857
Articles are hosted by Taylor and Francis Online.
The effects of 232U and 236U on uranium recycling in boiling water reactors are studied with the two-dimensional lattice physics code TGBLA. A simple analytic expression is proposed for reactivity compensation factor K, taking into account the self-shielding effect of resonance absorption in 236U: K = a + b/ (1 + ce6)1/2, where e6 denotes the 236U concentration. To output the same energy as the 3.0 wt% enrichment fuel free from 236U, the constants are determined to be a = 0.06, b = 0.23, and c = 1.9. The introduction of 1 ppb 232U increases the surface dose rate of the fuel assembly by 60% over the aged enriched natural uranium. Lead time is as important as cooling time in 232U production because of the presence of the chain that originates from the alpha decay of naturally occurring 234U. The natural uranium feed and the separative work requirement are evaluated on these bases, introducing typical recycling strategies, and it appears that uranium recycling saves 17 to 19% of the natural uranium but increases the separative work by 0 to 2%. The front-end cost analysis reveals the benefit of a concentrated utilization of reprocessed uranium, which results from the self-shielding effect of 236U and the assumption of a linear dependence of the front-end penalty on 232U concentration. Also studied are plutonium composition in irradiated fuels and the effects of extended burnup.