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Division Spotlight
Nuclear Installations Safety
Devoted specifically to the safety of nuclear installations and the health and safety of the public, this division seeks a better understanding of the role of safety in the design, construction and operation of nuclear installation facilities. The division also promotes engineering and scientific technology advancement associated with the safety of such facilities.
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.
K. Lassmann, Timm Preusser
Nuclear Technology | Volume 60 | Number 3 | March 1983 | Pages 406-419
Technical Paper | LWR Control Materials—I and II / Nuclear Fuel | doi.org/10.13182/NT83-A33127
Articles are hosted by Taylor and Francis Online.
An advanced method has been developed for the specific purpose of calculating temperatures in fuel element structural analysis. Fuel, cladding, coolant, and structural temperatures are treated by a single system of equations. Melting of the fuel and cladding and boiling of the coolant are included in the model. The method is compared to other solution techniques. The thermal characteristics of the finite element method (FEM) and finite difference method (FDM) transient calculations are compared. The present method includes FDM and FEM algorithms as special cases; an optimum combination of both techniques is the standard usage. Explicit, implicit, or Crank-Nicholson integration procedures are possible. The method is fast running, reliable, and has no stability problems. The new method has been implemented into the temperature calculation subcode system TEMPER for use with URANUS or other fuel element codes. Special attention has been given to user requirements (e.g., an automatic time-step control). The URANUS code, with this subcode system TEMPER, has been applied successfully to difficult fast breeder fuel rod analysis including transient overpower, loss of flow, local coolant blockage, and specific carbide fuel experiments.