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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.
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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.
Siyang Huang, Qiqi Yan, Wenxi Tian, G. H. Su, Suizheng Qiu (Xi’an Jiaotong Univ)
Proceedings | Advances in Thermal Hydraulics 2018 | Orlando, FL, November 11-15, 2018 | Pages 44-56
In the nuclear power system, the critical heat flux (CHF) plays a crucial role in the reactor safety analysis. When CHF occurs, it will cause a sudden increase in the surface temperature, which would lead to the failure of fuel claddings and damage of the core. Considering the cross flow between neighboring channels, spacer grids and mixing vanes in the fuel assembly, the local flow conditions and the geometry of the flow channels make the prediction of CHF more complicated. In this paper, the departure from nucleate boiling (DNB) type CHF in rod bundle is investigated based on the coupled analysis of the subchannel method and a CHF mechanism model, i.e. the liquid sublayer dryout model. The liquid sublayer dryout model assumes that there is a thin liquid sublayer underneath a vapor blanket formed by the coalescence of small bubbles near the heated wall. The dryout of this sublayer will be regarded as the CHF occurrence. In present research, the homogeneous flow model is adopted in the subchannel analysis code to predict the local flow conditions for the rod bundle flow subchannels, which will be used as the input parameters for the liquid sublayer dryout model. In order to verify the method above, the predicted results are compared with the CHF Look-Up Table 2006 (LUT-2006) and the predicted results are in good agreement with the data in LUT-2006. In addition, the effects of rod bundle inlet subcooling, mass flux, heated length and motion conditions on CHF are analyzed.