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Division Spotlight
Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
Meeting Spotlight
ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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February 2025
Latest News
Fabrication milestone for INL’s MARVEL microreactor
A team from Idaho National Laboratory and the Department of Energy’s Office of Nuclear Energy (DOE-NE) recently visited Carolina Fabricators Inc. (CFI), in West Columbia, S.C., to launch the fabrication process for the primary coolant system of the MARVEL microreactor. Battelle Energy Alliance (BEA), which manages INL, awarded the CFI contract in January.
Mukesh Tayal, Ed Mischkot, Harve E. Sills, A. W. L. Segel
Nuclear Technology | Volume 76 | Number 2 | February 1987 | Pages 209-220
Technical Paper | Nuclear Fuel | doi.org/10.13182/NT87-A33875
Articles are hosted by Taylor and Francis Online.
The ELOCA-A code models the thermomechanical behavior of CANDU fuel elements during high-temperature transients such as loss-of-coolant accidents. Calculations include sheath and pellet temperatures, strains (including creep), sheath oxidation, and beryllium-assisted cracking. The ELOCA-A code was developed by adding axial nodes to the ELOCA·MK2 code, which assumes axially uniform temperatures and strains. Thus, it is now possible to study the effects of axial variations such as end flux peaking, axial variations in the microstructure of Zircaloy due to brazing, axially nonuniform heat transfer, and axially nonuniform cross section due to the presence of appendages. Other features of ELOCA-A include choice of Urbanic-Heidrick or Baker-Just correlations for sheath oxidation and double-sided oxidation of a failed sheath. The ELOCA-A code shows reasonable agreement with axial variations in hoop strains measured at Chalk River Nuclear Laboratories. Calculations for some arbitrary transients confirm that axial variations in initial microstructure and in neutron flux can have a significant effect on fuel temperatures and strains.
