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Robotics & Remote Systems
The Mission of the Robotics and Remote Systems Division is to promote the development and application of immersive simulation, robotics, and remote systems for hazardous environments for the purpose of reducing hazardous exposure to individuals, reducing environmental hazards and reducing the cost of performing work.
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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 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.”
Akio Yamamoto, Tsutomu Ikeno
Nuclear Technology | Volume 149 | Number 2 | February 2005 | Pages 175-188
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT05-A3588
Articles are hosted by Taylor and Francis Online.
In this paper, the effect of a pin-by-pin thermal-hydraulic feedback treatment on the core characteristics at a steady-state condition is investigated using a three-dimensional fine-mesh core calculation code. Currently, advanced nodal codes treat the inside of an assembly as homogeneous, and the temperature distribution inside a node is usually ignored. Namely, the fuel temperature is estimated from the assembly average power density, and the moderator temperature is calculated from the nodewise closed-channel model. However, the validity of a flat temperature distribution inside a node has not yet been investigated, because a three-dimensional pin-by-pin whole-core calculation must be done for comparison. A three-dimensional pin-by-pin nodal-transport code for a pressurized water reactor (PWR) core analysis, SCOPE2, was used in this study since it can directly treat the pin-by-pin feedback effect. A whole-core subchannel analysis code was developed to enhance the thermal-hydraulic capability of SCOPE2. The pin-by-pin feedback models for fuel and moderator temperature were established, and their impact on the core characteristics was investigated in a 3 × 3 multiassembly and the whole PWR core geometries. The calculations showed that modeling of the pin-by-pin temperature distribution revealed a negligible effect on core reactivity and only a slight impact on the radial peaking factor. The difference in the radial peaking factor that is exposed by the pin-by-pin temperature modeling is less than 0.005 in the test calculations.