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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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Nuclear Science and Engineering
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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.”
Kazuyuki Takase, Yasuo Ose, Hajime Akimoto
Fusion Science and Technology | Volume 39 | Number 2 | March 2001 | Pages 1050-1055
Safety and Environment | doi.org/10.13182/FST01-A11963382
Articles are hosted by Taylor and Francis Online.
Damage of cooling tubes of plasma facing components (PFCs) results in water discharge into a vacuum vessel (W) of a fusion reactor. Flashing in vacuum, water pool boiling and impingement-jet on a surface of the PFC are the main heat transfer phenomena responsible for steam production that causes a rapid pressurization of the W. This is called an in-vessel loss-of-coolant accident (LOCA) event or ingress-of-coolant event (ICE). The ICE event is one of the most severe accidents in the fusion reactors.
The integrated ICE test facility was constructed to demonstrate the safety design approach of International Thermonuclear Experimental Reactor (ITER) and obtain validation data for the ITER safety analysis codes. Then, an experimental study was performed using the integrated ICE test facility and at the same time the code validation study with the TRAC code was carried out. The pressure rise characteristics in the current ITER machine during the ICE event were analyzed numerically using the verified TRAC-PF1 code and the effects of the relief pipe diameter and suppression tank volume regarding to the pressure rise due to the ICE events were clarified quantitatively from the present analytical results.