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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott Downtown
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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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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.
Kaushik Banerjee, Kevin R. Robb, Georgeta Radulescu, John M. Scaglione, John C. Wagner, Justin B. Clarity, Robert A. LeFebvre, Joshua L. Peterson
Nuclear Technology | Volume 195 | Number 2 | August 2016 | Pages 124-142
Technical Paper | doi.org/10.13182/NT15-112
Articles are hosted by Taylor and Francis Online.
A novel assessment has been completed to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing-basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance. These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed, calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014), and significant uncredited transportation dose rate margins were also observed. The results demonstrate that at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.