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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.
Meeting Spotlight
Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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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NRC engineers share their expertise at the University of Puerto Rico
Robert Roche-Rivera and Marcos Rolón-Acevedo are licensed professional engineers who work at the U.S. Nuclear Regulatory Commission. They are also alumni of the University of Puerto Rico–Mayagüez (UPRM) and have been sharing their knowledge and experience with students at their alma mater since last year, serving as adjunct professors in the university’s Department of Mechanical Engineering. During the 2023–2024 school year, they each taught two courses: Fundamentals of Nuclear Science and Engineering, and Nuclear Power Plant Engineering.
Workshop
Sunday, October 3, 2021|11:00AM–1:00PM EDT
Session Chair:
Dean Wang (The Ohio State Univ.)
Student Producer:
Khaldoon Al-Dawood (NC State Univ.)
Speaker: Dean Wang (The Ohio State University).
It has been well known that the analytic neutron transport solution limits to the analytic solution of a diffusion problem for optically thick systems with small absorption and source. The standard technique for proving the asymptotic diffusion limit is constructing an asymptotic power series of the neutron angular flux in small positive parameter, which is the ratio of a typical mean free path of a particle to a typical dimension of the domain under consideration. In this workshop, we will present a new proof to directly show that the analytical neutron transport solution satisfies the diffusion equation at the asymptotic limit based on a recently obtained closed-form analytical solution of the monoenergetic SN neutron transport equation in slab geometry. In numerical solution of the SN neutron transport equation, a spatial discretization is of practical interest if it possesses the optically thick diffusion limit. Such a numerical scheme will yield accurate solutions for diffusive problems if the spatial mesh size is thin with respect to a diffusion length, whereas the mesh cells are thick in terms of a mean free path. We will present a recently obtained theoretical result on the asymptotic diffusion limit of numerical schemes and what mesh sizes should be used to achieve accurate results. In addition, we will present an interesting implication of the asymptotic diffusion limit on Fourier analysis for CMFD schemes. Audience: anyone.
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