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Division Spotlight
Operations & Power
Members focus on the dissemination of knowledge and information in the area of power reactors with particular application to the production of electric power and process heat. The division sponsors meetings on the coverage of applied nuclear science and engineering as related to power plants, non-power reactors, and other nuclear facilities. It encourages and assists with the dissemination of knowledge pertinent to the safe and efficient operation of nuclear facilities through professional staff development, information exchange, and supporting the generation of viable solutions to current issues.
Meeting Spotlight
2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott Downtown
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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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.
Thomas E. Michener, David R. Rector, Judith M. Cuta
Nuclear Technology | Volume 199 | Number 3 | September 2017 | Pages 350-368
Technical Paper | doi.org/10.1080/00295450.2017.1327253
Articles are hosted by Taylor and Francis Online.
The COBRA-SFS thermal-hydraulic code has been incorporated into the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System tool as a module devoted to spent-fuel-package thermal analysis. COBRA-SFS has been extensively validated and widely applied to thermal-hydraulic analysis of a large range of spent-fuel storage systems. Instead of recapping that long and detailed history, this paper summarizes the most significant and unique verification and validation of COBRA-SFS, which consists of comparison of code temperature predictions to experimental data obtained in the Test Area North Facility at the Idaho National Laboratory in the 1980s and early 1990s. These data were obtained as part of a program undertaken by the U.S. Department of Energy Office of Civilian Radioactive Waste Management for thermal performance testing of commercial spent-fuel storage cask designs. In total, four casks were tested, and all tests were performed with Westinghouse 15×15 pressurized water reactor spent fuel from the Surry or Turkey Point reactors. COBRA-SFS code results and experimental data comparisons are shown only for the CASTOR-V/21 and the TN-24P casks. CASTOR-V/21 was loaded with the highest decay heat load tested in this program, with individual assembly decay heat values up to 1.83 kW. This effectively bounds storage conditions currently contemplated for high-heat-load systems with test conditions reaching fuel cladding temperatures that approached and in some cases exceeded 400°C, the current regulatory limit for peak cladding temperature in dry storage. TN-24P, with a decay heat load of 20.5 kW, provides comparisons with experimental data that represent a realistic upper bound on typical dry storage initial conditions in independent spent fuel storage installations around the country. The consistency and accuracy of the COBRA-SFS temperature predictions in comparison to the measured data from these casks show that the code appropriately predicts the thermal-hydraulic and heat transfer behavior of these systems. The results presented here provide an excellent illustration of the capability of the COBRA-SFS code to correctly capture all three modes of heat transfer (thermal radiation, conduction, and convection) and the internal circulation of the backfill gas within a spent-fuel package in horizontal or vertical orientation.