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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.
Meeting Spotlight
ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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
First astatine-labeled compound shipped in the U.S.
The Department of Energy’s National Isotope Development Center (NIDC) on March 31 announced the successful long-distance shipment in the United States of a biologically active compound labeled with the medical radioisotope astatine-211 (At-211). Because previous shipments have included only the “bare” isotope, the NIDC has described the development as “unleashing medical innovation.”
Akitoshi Hotta, Hiroshi Shirai, Shinya Mizokami
Nuclear Science and Engineering | Volume 152 | Number 3 | March 2006 | Pages 292-305
Technical Paper | doi.org/10.13182/NSE06-A2583
Articles are hosted by Taylor and Francis Online.
A postulated single control rod drop transient was calculated for a typical boiling water reactor plant taking into account effects of detailed void distributions in those bundles neighboring the withdrawn control blade. Time-dependent pin power distributions were reconstructed by the plant simulator TRAC/BF1-ENTRÉE and were exported to the subchannel code NASCA.Macroscopic cross-section libraries based on flat and distorted void distributions were allocated in accordance with fuel location in a simplified two-way coupling method. Exposure trends of bundle neutronic properties were compared between two void distributions. Although the infinite multiplication factor was not influenced, the radial peaking factor increased significantly because of the void distortion caused by pin-by-pin exposure of fissile materials.The result with the combined cross sections was compared with those with the flat void cross sections. Application of the combined cross sections lowered the initial local peaking because of larger neutron leakage around the withdrawn control blade. The transient linear power density at the critical fuel rod increased more rapidly. A change in the fuel heat flux was attenuated because of the heat conduction delay. As a consequence of these effects, the peak cladding temperature became slightly lower than that of the flat void model.