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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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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Hash Hashemian: Visionary leadership
As Dr. Hashem M. “Hash” Hashemian prepares to step into his term as President of the American Nuclear Society, he is clear that he wants to make the most of this unique moment.
A groundswell in public approval of nuclear is finding a home in growing governmental support that is backed by a tailwind of technological innovation. “Now is a good time to be in nuclear,” Hashemian said, as he explained the criticality of this moment and what he hoped to accomplish as president.
Philippe Planquart, Chiara Spaccapaniccia, Giacomo Alessi, Sophia Buckingham, Katrien Van Tichelen
Nuclear Technology | Volume 206 | Number 2 | February 2020 | Pages 231-241
Technical Paper | doi.org/10.1080/00295450.2019.1637240
Articles are hosted by Taylor and Francis Online.
The thermal-hydraulic challenges of a nuclear reactor are numerous and mastering them is crucial for the design and safety of new reactors. Numerical simulation through computational fluid dynamics (CFD) codes or system thermal-hydraulic codes can address a lot of the different questions, nevertheless the use of water modeling for the study of the thermal-hydraulic behavior of a new primary system and the validation of codes remains an extremely valuable tool. A water model of the pool-type PbBi-cooled MYRRHA reactor has been developed at the von Karman Institute in collaboration with SCK•CEN. It is a full plexiglass model at a geometrical scale 1/5 of MYRRHA. This transparent water model allows the application of optical measurement techniques like particle image velocimetry (PIV) for flow characterization. Local results of PIV measurements performed in the lower plenum at the entrance of the core are presented and compared with CFD results for nominal operating condition and a natural convection case simulating decay heat removal. Very good agreement has been found in the velocity field. The results also show the importance of the radial flow entering the core of the water model in natural convection.