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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
Meeting Spotlight
2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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ANS names new Fellows, award winners ahead of Annual Meeting
The American Nuclear Society extends its congratulations to the new ANS Fellows and recipients of this year’s honors and awards, which will be presented at the President’s Special Session during the opening plenary of the 2024 ANS Annual Conference. Those being honored this year have made outstanding contributions to nuclear science and technology. The full list of awards follows below.
William Boyd, Adam Nelson, Paul K. Romano, Samuel Shaner, Benoit Forget, Kord Smith
Nuclear Technology | Volume 205 | Number 7 | July 2019 | Pages 928-944
Regular Technical Paper | doi.org/10.1080/00295450.2019.1571828
Articles are hosted by Taylor and Francis Online.
High-fidelity deterministic transport codes require highly accurate multigroup cross sections (MGXS). Monte Carlo is increasingly cited as a reactor-agnostic approach to MGXS generation since it is unconstrained by the engineering-based approximations that limit the applicability of deterministic MGXS generation tools. This paper introduces a new framework that uses the OpenMC Monte Carlo code to generate MGXS for use in multigroup transport codes. The openmc.mgxs module is built atop OpenMC’s Python application programming interface to process tally data output by the OpenMC executable. This paper validates the module to generate MGXS that enable the multigroup OpenMOC transport code to compute eigenvalues to within 50 pcm and fission rates to within 1% of reference solutions for two heterogeneous pressurized water reactor benchmarks.