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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
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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Jul 2024
Jan 2024
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Nuclear Science and Engineering
August 2024
Nuclear Technology
Fusion Science and Technology
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.”
Faith R. Beck, Yue Jin, Fan-Bill Cheung (Penn State), Stephen M. Bajorek, Kirk Tien, Chris L. Hoxie (NRC)
Proceedings | Advances in Thermal Hydraulics 2018 | Orlando, FL, November 11-15, 2018 | Pages 739-750
In a large break loss of coolant accident (LOCA) in a pressurized water reactor (PWR), a break may occur on one of the cold legs. The emergency core cooling system (ECCS) must provide sufficient coolant to the core to remove decay heat and prevent the cladding from exceeding 1477.6 K. During reflood, flow to the core is most often gravity driven, resulting in an oscillatory delivery of coolant to the core. These oscillations are attributed to vapor generation in the core and the dynamic response of the downcomer water level. The amount of liquid entrained can be significant because in an accident scenario, the progress rate will be slowed and it can take longer to fully recover the core. Most reflood experiments have been conducted with constant forced reflood rates, and have not explored in detail the effects of oscillations on liquid entrainment and rod bundle heat transfer. The few studies that have been conducted for oscillating flows indicate enhanced entrainment of liquid at the quench front. At the NRC-PSU Rod Bundle Heat Transfer (RBHT) Test Facility, an electrically heated 7×7, 3.66m long rod bundle array has the capabilities to perform both constant and oscillatory forced flooding rate experiments. For varying oscillation periods and inlet subcoolings, these system parameters’ effect the liquid entrainment has been analyzed. Using a frequency analysis and signal processing technique, the percent liquid entrainment has been determined in one up- or down-stroke for oscillatory reflood. These findings, which are important to rod bundle heat transfer during reflood transients, have been explained based on the physics of the problem.