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Division Spotlight
Reactor Physics
The division's objectives are to promote the advancement of knowledge and understanding of the fundamental physical phenomena characterizing nuclear reactors and other nuclear systems. The division encourages research and disseminates information through meetings and publications. Areas of technical interest include nuclear data, particle interactions and transport, reactor and nuclear systems analysis, methods, design, validation and operating experience and standards. The Wigner Award heads the awards program.
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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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.”
André Bergeron, Daniel Caruge, Philippe Clément
Nuclear Technology | Volume 134 | Number 1 | April 2001 | Pages 71-83
Technical Paper | NURETH-9 | doi.org/10.13182/NT01-A3187
Articles are hosted by Taylor and Francis Online.
The physical validation compared with the hydraulic and two-phase flow experiments of the thermal-hydraulic FLICA-IV nuclear core computer code, in the case of a pressurized water reactor is presented. This three-dimensional two-phase flow code is devoted to steady state and transient thermal-hydraulic analysis of nuclear reactor cores. The four balance equations used by the code and the closure relationships are first presented. Then, the facilities employed for the code validation are described. They are the ones that use either laser velocimetry techniques in the case of hydraulic validation to measure accurately the flow field around rods or isokinetic sampling to carry out the qualities and the axial mass velocities at the outlet of a rod bundle in the case of two-phase flow validation. Comparisons between experimental and computed values are then presented for the axial flow blockage simulation, inlet assemblies flow mixing, axial flow spacer grid disturbance, and an outlet rod bundle map of qualities and axial mass velocities.