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Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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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.
Chenglong Wang, Kaichao Sun, Lin-wen Hu, Dalin Zhang, Wenxi Tian, Suizheng Qiu, G. H. Su
Nuclear Technology | Volume 198 | Number 1 | April 2017 | Pages 1-16
Technical Paper | doi.org/10.1080/00295450.2017.1294011
Articles are hosted by Taylor and Francis Online.
A transportable fluoride-salt-cooled high-temperature reactor (TFHR) design with 20-MW(thermal) rated power and 18-month fuel cycle is proposed for off-grid applications. One of the design goals of the compact reactor core is potential transport by truck, rail, or air. Full-core thermal-hydraulic analyses and improvements using three-dimensional computational fluid dynamics (CFD) were performed previously to demonstrate the feasibility of a TFHR design at a nominal power of 20 MW(thermal). In this paper, the best-estimate system code Reactor Excursion Leak Analysis Program (RELAP5-3D) is adopted to study the transient behavior of this TFHR design and the safety characteristics of the primary loop system during accident conditions. The modeling results of the steady state were verified using CFD results with consideration of radial heat conduction between heat transfer unit cells. Four most challenging accidents of anticipated transient without scram were analyzed, as well as parametric studies of some key factors. These accidents include unprotected reactivity insertion accident (URIA), unprotected loss of heat sink (ULOHS), unprotected loss of flow (ULOF), and a combination accident of ULOF and ULOHS. The results indicate that transient temperature limits are not exceeded during the most severe accidents. They indicate satisfactory transient performance of the TFHR design. The transient temperature limit of structure material Hastelloy N, based on embrittlement phenomena, poses the most limiting constraint due to the small temperature margin of about 20 K in the accident combination of ULOF and ULOHS. Overall, TFHR is a sound reactor design from a thermal-hydraulic viewpoint.