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Division Spotlight
Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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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.”
J. E. M. Saxby, Anil K. Prinja, M. D. Eaton
Nuclear Science and Engineering | Volume 189 | Number 1 | January 2018 | Pages 1-25
Technical Paper | doi.org/10.1080/00295639.2017.1367569
Articles are hosted by Taylor and Francis Online.
The time and phase-space dependent backward master equation is used to develop and numerically solve a coupled system of transport equations for the probability distribution of the neutron number in subregions of a spherically symmetric, reflected, subcritical plutonium sphere. The number distributions are computed for a single initial neutron injected into the assembly and localized in phase space as well as in the presence of a uniformly distributed spontaneous fission source in the fissile region. A standard multigroup, discrete ordinates scheme with second-order spatial and fully implicit time discretization proved sufficiently accurate for this application. The results presented show complex behaviors arising from the material interface and spectral effects due to neutron slowing down that cannot be encapsulated in a lumped model. Additionally, low-order spatial moments were computed both by averaging the number distributions of finite order and directly solving the transport equations for the moments using the same numerical scheme. While generally excellent agreement is observed between the two approaches, the truncation order has a noticeable effect on the accuracy of the higher moments that are computed using the number distributions.