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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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2024 ANS Annual Conference
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Las Vegas, NV|Mandalay Bay Resort and Casino
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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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Latest News
Excelsior University student section awarded community education grant
The American Nuclear Society Student Section at Excelsior University in Albany, N.Y., was awarded a $5,000 grant from the ANS Student Section Strategic Fund initiative for its program, Empowering Tomorrow’s Nuclear Innovators: A Collaborative Approach to Nuclear Technology Education and Awareness.
Robert K. Salko, William D. Pointer, Marc-Oliver Delchini, William L. Gurecky, Kevin T. Clarno, Stuart R. Salttery, Victor Petrov, Annalisa Manera
Nuclear Technology | Volume 205 | Number 12 | December 2019 | Pages 1697-1706
Technical Paper | doi.org/10.1080/00295450.2019.1585734
Articles are hosted by Taylor and Francis Online.
The Consortium for Advanced Simulation of Light Water Reactors is developing a core simulator capability known as the Virtual Environment for Reactor Applications (VERA) to address nuclear industry challenge problems such as crud-induced power shift (CIPS). The CTF thermal-hydraulic (T/H) subchannel code provides thermal feedback in the coupled neutronics, T/H, crud chemistry simulation that VERA performs. It has been discovered that the coarse meshing approach used by CTF (in which fuel rods are discretized into four azimuthal segments) can be a source of error in predicting crud growth and boron distribution in VERA CIPS calculations. Spacer grid effects lead to complex rod-to-fluid heat transfer behavior that, when not resolved, can lead to error in the prediction of crud growth and boron deposition. A higher-fidelity computational fluid dynamics approach can be used instead of CTF, but this leads to excessive simulation times. This paper presents an approach for using high-fidelity computational fluid dynamics data to create shape functions that are used in CTF to reconstruct rod surface heat transfer behavior as a function of spacer grid geometry. The approach is demonstrated for a 5 × 5 rod bundle facility with five mixing vane grids under a range of operating conditions encountered in nominal pressurized water reactor conditions. It is demonstrated that the grid heat transfer maps are successful at introducing a higher-fidelity heat transfer modeling capability into CTF.