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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott Downtown
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ANS announces 2025 Presidential Citations
One of the privileges of being president of the American Nuclear Society is awarding Presidential Citations to individuals who have demonstrated outstanding effort in some manner for the benefit of ANS or the nuclear community at large. Citations are conferred twice each year, at the Annual and Winter Meetings.
ANS President Lisa Marshall has named this season’s recipients, who will receive recognition at the upcoming Annual Conference in Chicago during the Special Session on Tuesday, June 17.
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.