ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
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Division Spotlight
Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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
Constellation seeks rezone for property adjacent to Illinois plant
While no development details have been released, Constellation is asking to rezone 658.8 acres of land it owns around the Byron nuclear plant in Illinois for possible long-term use.
M. Bradbury, R. Ratnayake (NuScale Power)
Proceedings | 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) | Charlotte, NC, April 8-11, 2018 | Pages 165-171
Data from Critical Heat Flux (CHF) and Critical Power (CP) tests are used for developing CHF and CP (or dryout) correlations to predict the maximum allowable heat flux at points of operational interest. CHF and CP tests are conducted using heater rod arrays representing fuel rods in a partial fuel bundle. Heat is directly generated in these heater rods depicting the axial power distributions of interest. The test rod array is housed in a slender vertical channel of metallic walls that are generally unheated. During testing, part of the heat generated in the heater rods is lost to the environment through the channel walls. Heat loss through the channel wall has the potential to favor channel thermal-hydraulic (T-H) conditions, and thereby make power measurements non-conservative. Typically, a single conservative estimate of the heat loss obtained from a separate isothermal test of the same test configuration is used regardless of the reactor conditions tested during a test campaign. This paper investigates the basis for the use of a single value for channel heat loss under varying reactor conditions. Results indicate that this practice has sound bases when applied with appropriate conservatisms.