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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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.
V. Marinelli, A. Pellei, P. Vallero, C. Vitanza
Nuclear Technology | Volume 27 | Number 2 | October 1975 | Pages 207-215
Technical Paper | Reactor | doi.org/10.13182/NT75-A24287
Articles are hosted by Taylor and Francis Online.
The (X, LB) correlation, which is currently used to predict the critical heat flux (CHF) onset in simple and complex geometries under a steady-state regime, is capable of a dynamic interpretation. The two-phase flow mean particles, climbing the channel at linear velocities corresponding to their mass velocity and local densities, reach the CHF conditions—i.e., zero thickness of the liquid film on the rod—when they have traveled a certain distance and have achieved a certain quality. According to this model, the CHF would be predicted in transient condition, when the boiling length and the mass velocities change with space and time, by applying the steady-state CHF (X, LB) correlation to the actual paths of the mean fluid particles. The calculations performed in comparison of the “Lagrangian point of view,” outlined above by means of the DOLCE computer code, with the local space-time approach of the “Eulerian point of view” indicate that the two methods give substantially equivalent results and predict satisfactorily the onset of the transient CHF for the Centro Informazioni Studi Esperienze annuli experimental data and General Electric Company 16-rod bundles data under typical boiling water reactor transients, including loss-of-coolant accident simulations.