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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
Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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
BWXT will scout potential TRISO fuel production sites in Wyoming
BWX Technologies Inc. announced today that its Advanced Technologies subsidiary has signed a cooperation agreement with the state of Wyoming to evaluate locations and requirements for siting a potential new TRISO nuclear fuel fabrication facility in the state.
Jason Chao, V. K. (Bindi) Chexal, William H. Layman, David A. Rautmann, Craig E. Peterson, Larry W. Cress
Nuclear Technology | Volume 61 | Number 2 | May 1983 | Pages 224-237
Technical Paper | Second International RETRAN Meeting / Heat Transfer and Fluid Flow | doi.org/10.13182/NT83-A33193
Articles are hosted by Taylor and Francis Online.
The RETRAN-02 and DYNODE-P thermal-hydraulic codes were compared against actual Prairie Island plant data from a steam generator tube break incident that occurred on October 2, 1979. The predictions from the code calculations compare well with actual plant behavior. The time of the break in the Prairie Island incident was found to be ∼260 s prior to scram with an initial break flow of 625 gal/min. Discharge coefficients are recommended for the calculations of critical flow from the break with extended Henry-Fauske and Moody critical flow models. In addition, a linear correlation was developed to predict the break flow with a given system depressurization rate for a Westinghouse two-loop plant.