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Division Spotlight
Operations & Power
Members focus on the dissemination of knowledge and information in the area of power reactors with particular application to the production of electric power and process heat. The division sponsors meetings on the coverage of applied nuclear science and engineering as related to power plants, non-power reactors, and other nuclear facilities. It encourages and assists with the dissemination of knowledge pertinent to the safe and efficient operation of nuclear facilities through professional staff development, information exchange, and supporting the generation of viable solutions to current issues.
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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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.
Nicholas T. Saltos,* Richard N. Christensen, Tunc Aldemir
Nuclear Technology | Volume 83 | Number 1 | October 1988 | Pages 93-109
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT88-A34178
Articles are hosted by Taylor and Francis Online.
A methodology is presented to determine the transient temperature distributions in fuel bundles under loss-of-coolant-accident (LOCA) conditions using a recently developed variational technique for the solution of radial-azimuthal heat conduction in the fuel rods and the modified view factor concept proposed by Uchida and Nakamura to model the radiative heat transfer between the rods. The variational technique is based on the Lebon-Labermont restricted variational principle and represents the temperature distribution in the rods at a given time during the LOCA via parabolic and circular trial functions in the radial and azimuthal directions, respectively. The methodology is implemented to a 4 × 4 boiling water reactor fuel bundle under typical LOCA conditions to investigate the effects of changes in rod heat transfer characteristics and simplifying modeling assumptions on predicted rod temperature distributions. The results show that these effects depend on the rod location in the assembly and LOCA phase under consideration and indicate that same degree of modeling detail may not be necessary for all the rods in the bundle at all times during the LOCA. An important advantage of the methodology is that it selects the optimum number of trial functions for each rod and for each time step in the simulation to reduce the computation time without compromising solution accuracy. Other advantages are that (a) the variational technique is faster than finite difference techniques for comparable accuracy and uses the same algorithm for one-dimensional radial and two-dimensional solutions, and (b) formulation of the heat conduction problem in the rods is compatible with the modular accident analysis codes already in use in the nuclear industry.