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Division Spotlight
Robotics & Remote Systems
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.
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.
H. Y. Yoon, I. K. Park, J. R. Lee, S. J. Lee, Y. J. Cho, S. J. Do, H. K. Cho, J. J. Jeong
Nuclear Science and Engineering | Volume 194 | Number 8 | August-September 2020 | Pages 633-649
Technical Paper | doi.org/10.1080/00295639.2020.1727698
Articles are hosted by Taylor and Francis Online.
A high-fidelity safety analysis method for pressurized water reactors (PWRs) is presented using a multiscale and multiphysics coupled code. Computational resolution of the conventional safety analysis can be greatly improved using this method in which the whole reactor vessel is modeled at a subchannel scale with around 5 million calculation meshes. Three-dimensional thermal hydraulics inside the reactor vessel is simulated using CUPID-RV with subchannel-scale thermal-hydraulic models for the reactor core. The subchannel models were validated using the legacy rod bundle experiments including single- and two-phase flow tests that were used in the validation of other subchannel analysis codes. The three-dimensional mesh was generated for the reactor vessel. Structured meshes were used in the core region for the subchannel model, and body-fitted unstructured meshes were applied for the downcomer, lower and upper plenums, and hot and cold legs. The number of meshes was optimized for a practical calculation. A three-dimensional core kinetics code (MASTER) and a one-dimensional system analysis code (MARS) were coupled with CUPID-RV for an accident analysis of PWRs. Subchannel-scale full-core steam line break accident analysis of the OPR1000 PWR was realized using the coupled code (MASTER/CUPID-RV/MARS) with a reasonable computation time, and thus, the present method can be used as a practical tool for three-dimensional safety analysis of PWRs.