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Division Spotlight
Robotics & Remote Systems
The Mission of the Robotics and Remote Systems Division is to promote the development and application of immersive simulation, robotics, and remote systems for hazardous environments for the purpose of reducing hazardous exposure to individuals, reducing environmental hazards and reducing the cost of performing work.
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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Latest News
Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
Y. Liao, K. Vierow
Nuclear Technology | Volume 152 | Number 3 | December 2005 | Pages 302-313
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT05-5
Articles are hosted by Taylor and Francis Online.
A pressurized water reactor steam generator tube rupture (SGTR) is of concern because it represents a bypass of the containment for radioactive materials to the environment. In a station blackout accident, tube integrity could be threatened by creep rupture, particularly if cracks are present in the tube walls. Methods are developed herein to improve assessment capabilities for SGTR by using the severe-accident code MELCOR. Best-estimate assumptions based on recent research and computational fluid dynamics calculations are applied in the MELCOR analysis to simulate two-dimensional natural circulation and to determine the relative creep-rupture timing in the reactor coolant pressure boundary components. A new method is developed to estimate the steam generator (SG) hottest tube wall temperature and the tube critical crack size for the SG tubes to fail first. The critical crack size for SG tubes to fail first is estimated to be 20% of the wall thickness larger than by a previous analysis. Sensitivity studies show that the failure sequence would change if some assumptions are modified. In particular, the uncertainty in the countercurrent flow limit model could reverse the failure sequence of the SG tubes and surge line.