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Division Spotlight
Reactor Physics
The division's objectives are to promote the advancement of knowledge and understanding of the fundamental physical phenomena characterizing nuclear reactors and other nuclear systems. The division encourages research and disseminates information through meetings and publications. Areas of technical interest include nuclear data, particle interactions and transport, reactor and nuclear systems analysis, methods, design, validation and operating experience and standards. The Wigner Award heads the awards program.
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 Science and Engineering | Volume 196 | Number 1 | January 2022 | Pages 75-97
Technical Paper | doi.org/10.1080/00295639.2021.1945393
Articles are hosted by Taylor and Francis Online.
The postulated dual-failure accident, i.e., loss of primary coolant flow along with impairment of the emergency coolant injection system, leads to peak fuel temperatures. It is well known that the temperature of the fuel assemblies is one of the significant factors that affect the outcome of an accident. Therefore, the present work aims to thoroughly investigate the thermal response of a single channel under postulated accident conditions. An experimental system was developed to capture the steady-state heat and temperature distribution in a representative 37-element fuel channel for a decay heat of 6.13 kW. Ohmic heating of the fuel rod simulators (FRSs) mimicked the generation of radioactive decay heat. Numerical simulation was also performed using the Fluent 19.1® code, and the discrete ordinates method was used to solve the radiative transfer equation. Based on the experimental results and the simulation results, it was found that the maximum Zircaloy-4 cladding temperature ≈850°C to 870°C was in the center ring. The temperature was found to vary around the circumference for each of the FRSs. Furthermore, the outer ring FRSs that had the lowest temperature developed the highest circumferential temperature gradient. In the pressure tube, the average circumferential temperature gradient obtained from the experiment and the simulation was 3.76°C/radian and 3.85°C/radian, respectively. Between the calandria tube and the moderator, the heat transfer coefficient was estimated to be around 822.3 W/m2‧K.