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Division Spotlight
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.
Meeting Spotlight
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
Standards Program
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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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.
S. Kliem, S. Danilin, A. Hämäläinen, J. Hádek, A. Keresztúri, P. Siltanen
Nuclear Science and Engineering | Volume 157 | Number 3 | November 2007 | Pages 280-298
Technical Paper | doi.org/10.13182/NSE07-A2728
Articles are hosted by Taylor and Francis Online.
Recently, three-dimensional neutron-kinetics core models have been coupled to advanced thermal-hydraulic system codes. These coupled codes can be used for the analysis of the whole reactor system. In the framework of the international association Atomic Energy Research (AER) on VVER Reactor Physics and Reactor Safety, two benchmarks for these code systems were defined. The reference reactor is the Russian VVER-440. The response of the reactor core to a symmetric and an asymmetric main steam line break should be investigated. So, different aspects of the coupling could be tested. As an additional feature, the participants had to use their own nuclear data.Each of these benchmarks was calculated by five different code systems. The comparison of the received solutions for the symmetric case shows good agreement in the evolution of the thermal hydraulics. When the core power reestablishes after recriticality, differences between the single solutions develop, mainly connected with the use of different nuclear data. Because of the increased complexity of the calculations, in the second benchmark differences between the thermal-hydraulic behavior in the single calculations were observed, additionally. These differences have their main origin in the behavior of the secondary side.The results of both benchmarks show the safety potential of the VVER-440 reactor. Even under very conservative conditions, no fuel rod failure was determined by the calculations, and the reactor was transferred into a subcritical final state.