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Operations & Power
Members focus on the dissemination of knowledge and information in the area of power reactors with particular application to the production of electric power and process heat. The division sponsors meetings on the coverage of applied nuclear science and engineering as related to power plants, non-power reactors, and other nuclear facilities. It encourages and assists with the dissemination of knowledge pertinent to the safe and efficient operation of nuclear facilities through professional staff development, information exchange, and supporting the generation of viable solutions to current issues.
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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.
Kenji Takeuchi, Michael Y. Young, Lawrence E. Hochreiter
Nuclear Science and Engineering | Volume 112 | Number 2 | October 1992 | Pages 170-180
Technical Paper | doi.org/10.13182/NSE112-170
Articles are hosted by Taylor and Francis Online.
Wallis’ flooding correlation is generalized for both small and large pipes by the use of the critical Kutateladze number. A drift flux correlation is then obtained that is tangential to the generalized flooding curve. A simple function of void fraction for the correlation parameter is sufficient to provide good agreement with steam generator test data, without using flow regime maps. After the drift flux correlation is determined with the large-pipe test, it is implemented in the TRAC-PD2 computer code to be tested against the flooding curve for a small-diameter pipe. The Chexal-Lellouche formulas are also applied to the data analysis, and the results are compared with the present correlations. Discussion is extended to the Zuber-Findlay method of data analyses for the drift velocity and the distribution parameter, in relation to the flooding curve.
