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Radiation Protection & Shielding
The Radiation Protection and Shielding Division is developing and promoting radiation protection and shielding aspects of nuclear science and technology — including interaction of nuclear radiation with materials and biological systems, instruments and techniques for the measurement of nuclear radiation fields, and radiation shield design and evaluation.
Meeting Spotlight
Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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
Vogtle-3 shuts down for valve issue
One of the new Vogtle units in Georgia was shut down unexpectedly on Monday last week for a valve issue that has since been investigated and repaired. According to multiple local news outlets, Georgia Power reported on July 17 that Unit 3 was back in service.
Southern Company spokesperson Jacob Hawkins confirmed that Vogtle-3 went off line at 9:25 p.m. local time on July 8 “due to lowering water levels in the steam generators caused by a valve issue on one of the three main feedwater pumps.”
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.