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Reactor Physics
The division's objectives are to promote the advancement of knowledge and understanding of the fundamental physical phenomena characterizing nuclear reactors and other nuclear systems. The division encourages research and disseminates information through meetings and publications. Areas of technical interest include nuclear data, particle interactions and transport, reactor and nuclear systems analysis, methods, design, validation and operating experience and standards. The Wigner Award heads the awards program.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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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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.
Mark J. Holowach, Lawrence E. Hochreiter, Fan-Bill Cheung, David L. Aumiller
Nuclear Technology | Volume 140 | Number 1 | October 2002 | Pages 18-27
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT02-A3320
Articles are hosted by Taylor and Francis Online.
Critical heat flux (CHF) at a low-flow condition in a small-hydraulic-diameter duct is an important phenomenon for a Materials Test Reactor/Advanced Test Reactor (MTR/ATR) design under a number of accident conditions, including reflood transients. Current CHF models in the literature, such as the Mishima/Nishihara and Oh/Englert CHF models, are based on macroscopic system parameters and not local thermal-hydraulic conditions. These macroscopic parameter-based models cannot be readily used for analysis in transient best-estimate thermal-hydraulic codes. The present work focuses on developing a low-flow-rate CHF correlation, based on local conditions, that is amenable to implementation into a best-estimate transient thermal-hydraulic code for a small-hydraulic-diameter duct. The model development proceeds with a means of correlating CHF data to local conditions parameters and then applying a correction factor to the resulting correlation, subsequently permitting accurate predictions over a range of pressures. An evaluation of the proposed local conditions-based CHF model is conducted by predicting independent sets of CHF experimental results over a range of flow rate, pressure, and subcooling conditions. Conclusions on the viability of the proposed CHF model and suggestions for future efforts in improving the reflood heat transfer CHF models for small-hydraulic-diameter ducts are provided with an evaluation of the model results.
