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Radiation Protection & Shielding
The Radiation Protection and Shielding Division is developing and promoting radiation protection and shielding aspects of nuclear science and technology — including interaction of nuclear radiation with materials and biological systems, instruments and techniques for the measurement of nuclear radiation fields, and radiation shield design and evaluation.
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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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Crash Course: The DOE’s Package Performance Demonstration
Inspired by a history of similar testing endeavors and recommended by the National Academy of Sciences and the Blue Ribbon Commission on America’s Nuclear Future, the Department of Energy is planning to conduct physical demonstrations on rail-sized spent nuclear fuel transportation casks. As part of the project, called the Spent Nuclear Fuel Package Performance Demonstration (PPD), the DOE is considering a number of demonstrations based on regulatory tests and realistic transportation scenarios, including collisions, drops, exposure to fire, and immersion in water.
Roger L. Martz
Nuclear Technology | Volume 180 | Number 3 | December 2012 | Pages 316-335
Technical Paper | Special Issue on the Initial Release of MCNP6 / Radiation Transport and Protection | doi.org/10.13182/NT12-A15347
Articles are hosted by Taylor and Francis Online.
Los Alamos National Laboratory Monte Carlo N-Particle transport code (MCNP) Version 6 (MCNP6) has been extended to include a new capability that permits tracking of neutrons and photons on an unstructured mesh that is embedded as a mesh universe within its constructive solid geometry capability. The mesh geometry is created through Abaqus/CAE using its solid modeling capabilities. MCNP transport results are calculated for mesh elements using a path length estimator while element-to-element tracking is performed on the mesh. The results from MCNP6 can be exported to Abaqus/CAE for visualization or other physics analysis. Three geometrically simple benchmark experiments were analyzed: Godiva, Osaka nickel sphere, and fusion neutron source vanadium cube. Computer run time is proportional to the number of mesh elements, element order, and element type specified in the input. Good agreement of our MCNP6 results with the measured neutron leakage for the nickel sphere and the measured neutron and gamma spectra from the vanadium assembly was observed.
