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The Mission of the Robotics and Remote Systems Division is to promote the development and application of immersive simulation, robotics, and remote systems for hazardous environments for the purpose of reducing hazardous exposure to individuals, reducing environmental hazards and reducing the cost of performing work.
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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott 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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ANS announces 2025 Presidential Citations
One of the privileges of being president of the American Nuclear Society is awarding Presidential Citations to individuals who have demonstrated outstanding effort in some manner for the benefit of ANS or the nuclear community at large. Citations are conferred twice each year, at the Annual and Winter Meetings.
ANS President Lisa Marshall has named this season’s recipients, who will receive recognition at the upcoming Annual Conference in Chicago during the Special Session on Tuesday, June 17.
S. W. Hong, Y. S. Na, S. H. Hong, J. H. Song
Nuclear Technology | Volume 196 | Number 3 | December 2016 | Pages 538-552
Technical Paper | doi.org/10.13182/NT16-9
Articles are hosted by Taylor and Francis Online.
Some advanced reactors adapt the in-vessel corium retention concept by cooing the outside wall of the reactor vessel in severe accidents. If a reactor vessel failure happens in this case, the molten corium in the reactor vessel is directly injected into the water in the reactor cavity without the process of a free fall. Experiments using ZrO2 and molten corium to simulate the conditions in which the reactor vessel is fully flooded were recently carried out at the Test for Real cOrium Interaction with water (TROI) experimental facility, and the results are compared with the data produced under conditions in which the reactor vessel is partially flooded. It was observed that the melt front velocity in the water under submerged reactor conditions is much faster than that under partially flooded reactor cavity conditions, and a large bubble was observed at the surface of the mixing zone under submerged reactor conditions. Accordingly, it is estimated that the breakup of the melt jet in the water during the fuel-coolant interaction (FCI) test under submerged reactor conditions would be different than that of the FCI test under partially flooded reactor cavity conditions.