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Reactor Physics
The division's objectives are to promote the advancement of knowledge and understanding of the fundamental physical phenomena characterizing nuclear reactors and other nuclear systems. The division encourages research and disseminates information through meetings and publications. Areas of technical interest include nuclear data, particle interactions and transport, reactor and nuclear systems analysis, methods, design, validation and operating experience and standards. The Wigner Award heads the awards program.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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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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.”
R. Wayne Houston
Nuclear Science and Engineering | Volume 4 | Number 2 | August 1958 | Pages 227-238
doi.org/10.13182/NSE58-A15364
Articles are hosted by Taylor and Francis Online.
For samples exposed in high neutron flux regions of reactors the contribution to the total dosage due to the recoils from elastically scattered fast neutrons may be significant. The calculation of this contribution is considered here. Three methods are presented, each differing in the manner in which the details of the energy distribution of fast neutrons are treated. In the first, the neutron flux per unit energy interval is assumed to be of the asymptotic or 1/E form up to fission energies. In the second and third, a separate computation is made for the uncollided neutrons reaching the sample. The remaining contribution due to once-scattered neutrons is treated as in the first method, but alternate forms for the source spectrum of once-scattered neutrons are considered. Use of the equations requires only a knowledge of the thermal neutron flux in the vicinity of the sample. Assumptions and limitations are discussed. Numerical results are presented for comparison of the effects in light water, heavy water, and graphite moderated reactors in the irradiation of a hydrocarbon (cyclohexane) sample.