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Division Spotlight
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.
Meeting Spotlight
2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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
Commercial nuclear innovation "new space" age
In early 2006, a start-up company launched a small rocket from a tiny island in the Pacific. It exploded, showering the island with debris. A year later, a second launch attempt sent a rocket to space but failed to make orbit, burning up in the atmosphere. Another year brought a third attempt—and a third failure. The following month, in September 2008, the company used the last of its funds to launch a fourth rocket. It reached orbit, making history as the first privately funded liquid-fueled rocket to do so.
Sule Ergun, Jason G. Williams, Lawrence E. Hochreiter, Hergen Wiersema, Marcel Slootman, Marek Stempniewicz
Nuclear Technology | Volume 156 | Number 1 | October 2006 | Pages 69-74
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT06-A3774
Articles are hosted by Taylor and Francis Online.
Critical heat flux (CHF) at a natural boiling condition is an important phenomenon for a research reactor having a small-hydraulic-diameter geometry under a large-break loss-of-coolant accident condition. Accurately predicting the CHF under this condition is very important; therefore, the CHF models used in the best-estimate codes must be validated using appropriate experimental data for a given geometry. The present work focuses on validating the CHF calculations and models within the COolant Boiling in Rod Arrays-Two Fluid (COBRA-TF) code by simulating two sets of experiments, which were performed in tubes and annuli with different length-to-diameter ratios. In this work, the cocurrent upflow and downflow correlations developed by Mishima and Nishihara and Holowach et al. and Zuber correlations for the CHF used in COBRA-TF are validated against the experimental data obtained by Monde and Yamaji and Islam et al. Conclusions on the predictive capability of COBRA-TF for the CHF calculations for small-hydraulic-diameter geometry under natural boiling conditions are provided with the description of the correlations and models used.