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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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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.”
Satoshi Suzuki, Kazuyoshi Sato, Masanori Araki, Kazuyuki Nakamura, Masayuki Dairaku, Kenji Yokoyama, Masato Akiba
Fusion Science and Technology | Volume 30 | Number 3 | December 1996 | Pages 788-792
Plasma-Facing Components: Analysis and Technology | doi.org/10.13182/FST96-A11963032
Articles are hosted by Taylor and Francis Online.
The divertor plate of next generation tokamak device such as the International Thermonuclear Experimental Reactor (ITER) is subjected to high heat and particle fluxes from the plasma. The armor material of the divertor plate will be damaged by plasma disruptions. Therefore the divertor components are required to be easily repaired or replaced. To realize the easy maintenance of divertor high heat flux components, the thermal bond layer (TBL) concept is the most promising. The TBL is a soft braze material which has a low melting temperature such as lead-based alloys. By using TBL for divertor high heat flux components, the damaged armor tiles are expected to be easily replaceable. Moreover, TBL plays a role of a compliant interlayer to reduce the interfacial thermal stress between the armor tile and the heat sink substrate. The authors have developed and tested Divertor Mock-UPS with lead-based TBL. This paper presents the results of the high heat flux experiments on the Divertor Mock-UPS with the TBL. The mock-up consists of replaceable armor blocks and a permanent substrate which is made of pure copper. The armor block consists of a bonded structure with armor tiles and a copper pivot. The armor blocks are brazed onto the substrate with pure lead. The permanent substrate has parallel cooling tubes which have twisted tape insert to enhance heat transfer. The high heat flux experiments on these mock-ups were carried out at JAERI Electron Beam Irradiation System (JEBIS). For the simulation of normal operation of ITER, the incident heat flux of 5 MW/m2 at a maximum pulse duration of 50 s was loaded; and for the simulation of transient period of ITER, the heat flux of up to 15 MW/m2 at a maximum pulse duration of 10 s was loaded. The thermal performance of the mock-ups was stably sustained against both thermal loads.