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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver Downtown
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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
Argonne’s METL gears up to test more sodium fast reactor components
Argonne National Laboratory has successfully swapped out an aging cold trap in the sodium test loop called METL (Mechanisms Engineering Test Loop), the Department of Energy announced April 23. The upgrade is the first of its kind in the United States in more than 30 years, according to the DOE, and will help test components and operations for the sodium-cooled fast reactors being developed now.
Charles T. Kelsey IV, Anil K. Prinja
Nuclear Technology | Volume 168 | Number 2 | November 2009 | Pages 257-263
Neutron Data | Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection | doi.org/10.13182/NT09-A9191
Articles are hosted by Taylor and Francis Online.
The limited availability of coupled multigroup proton/neutron cross-section libraries has hampered the use of deterministic transport methods for solving shielding problems involving energetic proton sources. Libraries are developed from evaluated nuclear data for low-energy transport and the physics models of MCNPX for intermediate-energy transport. They allow deterministic solutions of orbiting spacecraft shielding problems. Evaluated cross sections for protons and neutrons are available for many nuclides up to 150 MeV. NJOY99 is used to produce coupled multigroup proton/neutron cross sections from these. For higher energies, MCNPX is run in its cross-section calculation mode where the XSEX3 program is used to tally double-differential cross sections. The XSEX3 program was modified to discretize the cross sections in energy and output Legendre expansions for angular dependence. The NJOY99 and modified XSEX3 output are combined to produce cross-section libraries for energies up to 400 MeV. The libraries are used to solve trapped proton flux shielding problems using the discrete ordinates transport code Attila. High-order Legendre expansions (P39) are required to accurately describe the highly anisotropic scattering. Attila applies the extended transport correction allowing accurate three-dimensional solutions at much lower degrees. Particle flux solutions for orbiting spacecraft shielding problems obtained with Attila and MCNPX compare favorably. Coupled multigroup proton/neutron cross-section libraries, for use with deterministic transport codes, can be prepared using NJOY99 and MCNPX. Our results using the Attila code demonstrate that multigroup deterministic methods are computationally efficient alternatives to Monte Carlo simulation.