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Conference Spotlight
2025 ANS Winter Conference & Expo
November 9–12, 2025
Washington, DC|Washington Hilton
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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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Leading the charge: INL’s role in advancing HALEU production
Idaho National Laboratory is playing a key role in helping the U.S. Department of Energy meet near-term needs by recovering HALEU from federal inventories, providing critical support to help lay the foundation for a future commercial HALEU supply chain. INL also supports coordination of broader DOE efforts, from material recovery at the Savannah River Site in South Carolina to commercial enrichment initiatives.
Jun Yang, Michael Scott Greenwood, Matthew De Angelis, Michael Avery, Mark Anderson, Michael Corradini, James Matos, Floyd Dunn, Earl Feldman
Nuclear Science and Engineering | Volume 180 | Number 2 | June 2015 | Pages 141-153
Technical Paper | doi.org/10.13182/NSE14-45
Articles are hosted by Taylor and Francis Online.
A critical heat flux (CHF) experimental study at low pressure and natural convection condition has been conducted. The test apparatus is a natural circulation loop with an upward flow channel, simulating TRIGA (Training, Research, Isotopes, General Atomics) reactors. CHF is studied in three types of geometries: a single-rod annulus, a three-rod bundle in a trefoil tube, and a four-rod bundle in a square tube. The full-scale fuel pin heater rod is electrically heated with a prototypic axial power profile, equipped with thermocouples for CHF detection. Experiments are carried out at the following conditions: inlet subcooling from 10 to 70 K, pressure from 110 to 290 kPa, and mass flux from 0 to 400 kg/m2·s. It is observed that CHF increases as the pressure or mass flux increases but does not significantly depend on the inlet subcooling within the testing range. The current CHF data are compared with a few selected CHF correlations whose application ranges are close to the testing conditions. The relevance of the CHF to the testing parameters is investigated. A modified CHF correlation compatible with TRIGA reactor conditions is proposed based on a previous correlation and current experimental data.
