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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.
Sang Woon Shin, Hee Cheon No
Nuclear Technology | Volume 73 | Number 3 | June 1986 | Pages 378-383
Technical Paper | Material | doi.org/10.13182/NT86-A16079
Articles are hosted by Taylor and Francis Online.
To investigate the denting phenomenon, the rates of corrosion occurring in simulated tube/support plate crevices were examined by using seven model boilers at 1.11 MPa. The model boilers were operated with all-volatile treatment (morpholine + hydrazine) with 15- and 100-ppm chloride concentrations constituting 10% FeCl2, 30% NaCl, and 60% CaCl2. It was found that corrosion rates increased with heat flux. A model was proposed to explain this observation, based on mechanisms that acid chloride is concentrated in the tube/support plate crevices. The model is expressed by the following equation for empty heated crevices: Good agreement was obtained by comparing the results predicted by the model with Brown’s data and the present data for empty heated crevices, and with Pathania’s data obtained at high heat flux. Based on the above equation, a model was developed to describe chloride concentration within the crevices versus heat flux for given condenser leakage rates in nuclear steam generators. Results predicted by the model show that a small increase in condenser leakage rates gives a considerable increase in chloride concentration within the crevices.