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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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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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Latest News
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.
T. K. Mau, T. B. Kaiser, A. A. Grossman, A. R. Raffray, X. R. Wang, J. F. Lyon, R. Maingi, L. P. Ku, M. C. Zarnstorff, ARIES-CS Team
Fusion Science and Technology | Volume 54 | Number 3 | October 2008 | Pages 771-786
Technical Paper | Aries-Cs Special Issue | doi.org/10.13182/FST08-27
Articles are hosted by Taylor and Francis Online.
The critical issue of divertor configuration for heat and particle flux control in a conceptual ARIES compact stellarator (CS) reactor is addressed. The goal is to determine a divertor location and geometry with a peak heat load of not more than 10 MW/m2 for a CS equilibrium based on the configuration to be used in the NCSX experiment, optimized for high beta (6.4%) and designed for low alpha-particle power loss fraction (5%). The surface heat flux on the target has three components: thermal particles, lost energetic alphas, and radiation from the core and the scrape-off layer. The first two components are dominant and their magnitudes can be comparable. To maintain a tritium-breeding ratio of 1.1, the total target area should not exceed 15% of the boundary plasma surface area. The divertor concept consists of two pairs of target plates per field period, one pair each at the top and bottom of the plasma. The heat flux profile is assessed by assuming that the parallel transport can be represented by field line mapping and that cross-field transport can be modeled with a prescribed field line diffusion scheme. In this manner, the poloidal and toroidal extents of the plates and their shape and distance to the plasma are designed to intercept all the heat flux and to minimize the peak thermal heat load. An approximate scheme, based on particle drift orbits in the core and field line tracing in the edge, is derived to estimate the alpha-particle heat load distribution over the plates and the first wall. The best plate configuration to date yields total peak heat loads (thermal + alpha) ranging from 5 to 18 MW/m2. Further optimization of the target plates is required to reach the design goal, which will be addressed in a future study.