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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 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.”
Constantine P. Tzanos, B. Dionne
Nuclear Technology | Volume 179 | Number 3 | September 2012 | Pages 382-391
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT12-A14170
Articles are hosted by Taylor and Francis Online.
To support the safety analysis of the conversion of the BR2 research reactor to low-enriched uranium (LEU) fuel and extend the validation basis of the RELAP code for the analysis of the conversion of research reactors from highly enriched (HEU) fuel to LEU, the simulation of BR2 tests A/400/1, C/600/3, and F/400/1 was undertaken. These tests are characterized by loss of flow initiated at different reactor power levels with or without loss of system pressure, reactor scram, flow reversal, and reactor cooling by natural circulation. This work presents the RELAP analysis of tests C/600/3 and F/400/1 and comparison of code predictions with experimental measurements for peak cladding temperatures during the transient at different axial locations in an instrumented fuel assembly. The simulations show that accurate representation of the power distribution, especially after reactor scram, between the fuel assemblies and the moderator/reflector regions is critical for the correct prediction of the peak cladding temperatures during the transient. Detailed MCNP and ORIGEN simulations were performed to compute the power distribution between the fuel assemblies and the moderator/reflector regions. With these distributions the predicted peak cladding temperatures are in good agreement with experimental measurements.