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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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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.
R. Paviotti Corcuera
Nuclear Science and Engineering | Volume 58 | Number 3 | November 1975 | Pages 278-290
Technical Paper | doi.org/10.13182/NSE75-A26777
Articles are hosted by Taylor and Francis Online.
Systematic discrepancies between experimental and calculated neutron spectra in different fast-neutron media with high 238U content were observed, allowing the uncertainties in certain 238U neutron cross sections to be considered as the source of such discrepancies. Sensitivity studies, as well as the effect of the former uncertainties on such fast-neutron spectra showed that 238U (n, n′) inelastic scattering is by far the main parameter determining the spectrum over a large energy range. Uranium-238 (n, γ) neutron capture is important only at low energies, below the 238U (n, n′) threshold. The (n,γn′) reaction in 238U using Fricke's data leads to an increase in the discrepancies in neutron spectra though this cross section could be considered negligible according to other arguments. A general least-squares adjustment of the 238U total inelastic cross sections, on the neutron spectrum discrepancies, was carried out between 70 keV and 3.6 MeV, using two different inelastic probability matrices (those of ENDF/B-III and UKNDL-DFN 401), which were chosen among several evaluations of this reaction. The adjusted cross sections obtained in an energy mesh corresponding to lethargy intervals of 0.5 imply a systematic reduction of 20 to 30% relative to the initial values in both files, but they appear to be in good agreement with other recent (coarse-mesh) adjusted data, including the KFK-INR set.