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Human Factors, Instrumentation & Controls
Improving task performance, system reliability, system and personnel safety, efficiency, and effectiveness are the division's main objectives. Its major areas of interest include task design, procedures, training, instrument and control layout and placement, stress control, anthropometrics, psychological input, and motivation.
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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The 2025 ANS election results are in!
Spring marks the passing of the torch for American Nuclear Society leadership. During this election cycle, ANS members voted for the newest vice president/president-elect, treasurer, and six board of director positions (four U.S., one non-U.S., one student). New professional division leadership was also decided on in this election, which opened February 25 and closed April 15. About 21 percent of eligible members of the Society voted—a similar turnout to last year.
Rodolfo Vaghetto, Timothy Crook, Alessandro Vanni, Yassin A. Hassan
Nuclear Technology | Volume 193 | Number 1 | January 2016 | Pages 88-95
Technical Paper | Special Issue on the RELAP5-3D Computer Code | doi.org/10.13182/NT14-147
Articles are hosted by Taylor and Francis Online.
During a loss-of-coolant accident (LOCA), fibrous debris and other particles generated by the jet impingement may be transported to the sump, accumulate, or even penetrate through the strainers, reaching the reactor core. Pressure relief holes and other plant-specific features may provide alternative paths to the coolant under debris-generated core blockage scenarios and can play a major role in core coolability. A typical four-loop pressurized water reactor was modeled using RELAP5-3D to simulate the reactor system response during large-break LOCA scenarios under hypothetical full core blockage conditions. Pressure relief holes were included in the input model to study the effects of these alternative flow paths on the core coolability. The comparison of the simulation results obtained with two different models (with and without pressure relief holes) proved the effectiveness of these alternative flow paths in providing sufficient flow to the core to remove the decay heat during the long-term cooling phase, maintaining the cladding temperature sufficiently below the safety limits at any time after the core blockage occurred. The results presented in this paper not only confirmed the importance of including specific geometric features of the reactor system (generally neglected) when simulating core blockage scenarios but also provided evidence that even under certain extreme core blockage conditions, core coolability may still be guaranteed.