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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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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
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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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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.
L. W. Ward
Nuclear Technology | Volume 131 | Number 1 | July 2000 | Pages 69-81
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT00-A3105
Articles are hosted by Taylor and Francis Online.
A model was developed to compute the two-dimensional velocity profiles in hot fuel channels of a pressurized water reactor core following a small-break loss-of-coolant accident (SBLOCA). Following an SBLOCA, the transient two-phase level in the core recedes below the top of the core, exposing the core to steam cooling and heatup of the fuel. To compute the velocity distributions, the Navier-Stokes equations were solved in vorticity form using an explicit upwind finite difference numerical scheme. The model was applied to the well-known lid-driven cavity problem and the data in the literature for vertically heated channels. Comparison of the model to the data in the literature provided validation of the approach.Application of the model to the conditions at the time of the peak clad temperature during core uncovery for a typical limiting small cold-leg break in a pressurized water reactor further revealed that the hot-channel steam flow can vary dramatically at the hot spot due to the severe distortion in the axial steam flow that is characteristic of asymmetrically heated channels. The results of the evaluation support the need for a thorough technical basis for the steam flow rates that are typically assumed to cool the hot rods in many commercial fuel rod heatup codes. These codes typically assume a constant mass flow along the axis of the fuel rod to compute the cladding temperature response. Mixed convection is shown to reduce the channel average velocity along the axis of the fuel rod by as much as 15%. The reductions in channel velocity will produce an attendant increase in the peak clad temperature achieved during an SBLOCA. The results of this study suggest that for the steam velocities used to cool hot rods during an SBLOCA, one needs to consider the mixed-convection behavior that can affect the convective heat transfer in the upper portions of exposed nuclear fuel bundles.