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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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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 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. 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.”
Yong-Sik Yang, Yang-Hyun Koo, Dae-Ho Kim, Je-Geon Bang, Young-Woo Rhee, Dong-Joo Kim, Keon-Sik Kim, Kun-Woo Song
Nuclear Technology | Volume 178 | Number 3 | June 2012 | Pages 267-279
Technical Paper | Fuel Cycle and Management | doi.org/10.13182/NT12-A13593
Articles are hosted by Taylor and Francis Online.
This paper presents some of the key technologies in the area of fuel performance that Korea Atomic Energy Research Institute (KAERI) has developed for a dual-cooled annular fuel, which should be available before the annular fuel can be considered to be used in a commercial nuclear power plant. First, considering the characteristics of the annular fuel - that it has two coolant channels, outer and inner, and also two gaps between the pellet and cladding - KAERI has developed a computer code DUOS that calculates temperature, swelling, densification, and stress and strain in the annular fuel. The DUOS code was verified by comparing it with either ABAQUS or analytical solutions. The first irradiation test of sintered annular fuel pellets with different initial densities was performed in the HANARO reactor up to a pellet burnup of 10.9 MWd/kg U and then subjected to postirradiation examination. Gamma scanning along the axial direction of the irradiated fuel rods showed the geometrical integrity of the annular fuel pellets, ruling out the possibility that fragmented annular pellet cracks could move down along the axial direction of the fuel rod and hence the pellet stack length could be reduced. Macroscopy of the annular fuel pellets revealed many radial and circumferential cracks that could lead to different outer and inner gap sizes along the axial direction of the annular fuel rod, which would suggest that heat transfer to both the outer and inner coolant channels during the irradiation of annular fuel rods would depend on the axial profile of the two gaps along the axial direction. The swelling rate derived from density measurement of the annular fuel pellets with 98.0% theoretical density was 0.25 to 0.60 vol % per 10 MWd/kg U, corresponding to the one observed for solid fuel pellets irradiated at low temperature.