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Mathematics & Computation
Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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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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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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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.
Bahram Nassersharif, James S. Peery, Evelyn M. Mullen, Stephen R. Behling
Nuclear Technology | Volume 94 | Number 1 | April 1991 | Pages 28-43
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT91-A16219
Articles are hosted by Taylor and Francis Online.
This study evaluates two significantly different models of a Westinghouse 414 reactor system using the TRAC-PF1/MOD1 computer code for a small-break loss-of-coolant accident (SBLOCA). A coarse threedimensional model of the reactor vessel is developed. In the coarse model, three of the four reactor coolant loops are combined into one loop. A detailed three-dimensional model of the reactor vessel is also developed. In the detailed model, each of the four coolant loops is modeled separately. Both models are run to steady-state convergence until the calculated system parameters are in good agreement. In addition, the steady-state results of both models closely match operational parameters given in the final safety analysis report. From the self-consistent steady-state conditions, a 60-s transient calculation is performed with each model. The transient simulates a 4-in. SBLOCA. The overall results of code predictions for the two models closely agree, and the vessel global parameters for the two models are also in good agreement. However, the computer times for the two calculations are significantly different. The detailed model provides additional information that is unavailable with the less detailed model, such as temperature and void fraction distributions throughout different regions of the vessel. During the 60-s transient, the upper head in the detailed model shows extensive voiding. The upper head in the coarse model also shows voiding; however, the extent and exact location of the voiding are not available in the coarse model. During this transient, the core region does not show extensive voiding; however, the detailed model shows some localized boiling. The results indicate that the coarse model is sufficient for 4-in. SBLOCA studies. The computer time associated with TRAC-PF1/MOD1 calculation of the extremely detailed model is ∼100 times longer than the coarse model.