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Division Spotlight
Operations & Power
Members focus on the dissemination of knowledge and information in the area of power reactors with particular application to the production of electric power and process heat. The division sponsors meetings on the coverage of applied nuclear science and engineering as related to power plants, non-power reactors, and other nuclear facilities. It encourages and assists with the dissemination of knowledge pertinent to the safe and efficient operation of nuclear facilities through professional staff development, information exchange, and supporting the generation of viable solutions to current issues.
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
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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.”
Alp Tezbasaran, Maria N. Avramova, Kostadin N. Ivanov (NCSU), Osman S. Celikten (Hacettepe Univ)
Proceedings | Advances in Thermal Hydraulics 2018 | Orlando, FL, November 11-15, 2018 | Pages 729-738
In this work, the sub-channel thermal-hydraulic code CTF is applied to the hottest fuel assembly of a VVER-1000 core, aiming to investigate the code sensitivity to uncertainties of the initial and boundary conditions. The core thermal-hydraulic solver CTF is a modernized version of the COBRA-TF sub-channel code, which is being maintained and developed by the Reactor Dynamics and Fuel Modeling Group (RDFMG) at North Carolina State University (NCSU) in cooperation with Oak Ridge National Laboratory (ORNL).
In this study, first, a full core model of a VVER-1000 reactor with its initial loading pattern is created for the Monte Carlo neutronics code MCNP6 under normal operating conditions using ENDF/B VII.1 / NJOY99. The assembly power factors and the pin-powers of the hottest fuel assembly, obtained by MCNP6, are used as power boundary conditions in CTF. The hottest assembly is simulated to calculate the fuel, cladding, and coolant temperatures at normal operating conditions.
Uncertainty analyses are performed using Dakota 6.5 and it is observed that CTF predictions of fuel, cladding, and coolant temperatures are most sensitive to uncertainties in core average power and inlet coolant temperature.