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Division Spotlight
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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Latest News
Argonne’s METL gears up to test more sodium fast reactor components
Argonne National Laboratory has successfully swapped out an aging cold trap in the sodium test loop called METL (Mechanisms Engineering Test Loop), the Department of Energy announced April 23. The upgrade is the first of its kind in the United States in more than 30 years, according to the DOE, and will help test components and operations for the sodium-cooled fast reactors being developed now.
Ketan Ajay, Ravi Kumar, Akhilesh Gupta
Nuclear Technology | Volume 210 | Number 3 | March 2024 | Pages 457-470
Research Article | doi.org/10.1080/00295450.2023.2229190
Articles are hosted by Taylor and Francis Online.
A reactor core overheats due to decay heat generated in the fuel when an effective cooling medium is unavailable, such as in a loss-of-coolant accident combined with a loss of emergency core coolant. If the heat generated is not effectively dissipated, then at extreme temperatures, the structural strength of the bundle assembly may deteriorate, leading to slumping of fuel elements onto the inner wall of the pressure tube. It is essential to examine the temperature behavior of the channel containing fuel pins in a disassembled state in order to comprehend the impact of further thermally induced deformations in the channel during postulated accident conditions. Capturing the temperature of channel components at each circumferential position from experiments is extremely difficult; thus, a modeling tool is necessary to obtain a thorough circumferential temperature profile. This paper presents a numerical study that aims to study the temperature distributions in a 1-m-long pressurized heavy water reactor (PHWR) channel containing a disassembled fuel bundle. The channel geometry and the boundary conditions implemented were obtained from the experiment. A temperature profile for each channel element at every circumferential and axial location was obtained. A thorough comparison of the predicted and the reported experimental values was performed, and it was found that the predicted temperature behavior of the channel was consistent with the experimental data. Further simulations with different fuel element configurations and decay powers may be carried out; in addition, the results obtained may be used for coupled thermal-mechanical and thermal-mechanical-chemical simulations.