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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
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
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
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.
Bryan Bednarz, Bin Han, X. George Xu
Nuclear Technology | Volume 168 | Number 2 | November 2009 | Pages 270-273
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 Biology and Medicine | doi.org/10.13182/NT09-A9193
Articles are hosted by Taylor and Francis Online.
During radiation therapy treatments, neutron contamination can be a source of unwanted radiation dose to the patient and medical personnel. Accurate cross-section data is needed to characterize the neutron contamination in medical accelerators using Monte Carlo methods. In this study, a comparison of the photoneutron yields using the default LA150U and the Chinese Nuclear Data Center (CNDC) photonuclear cross sections was performed. Thick tungsten plates, each of 0.125-cm thickness (one-third radiation length), were directly irradiated by an electron beam in MCNPX. In order to match typical radiation therapy energy ranges, the energy distribution of the electron beam was modeled as a Gaussian distribution with a mean energy of 18.3 MeV and a 3% full-width at half-maximum. The photoneutron yield using the LA150U is consistently [approximately]12 to 17% higher than those from the CNDC data for each target thickness. The average photoneutron energy difference between the two cross-section libraries ranged from 3 to 42%. No major differences were seen between relative neutron fluences per solid angle for the two cross-section libraries. The discrepancies between the datasets provided above can be attributed to the oversimplification of using the default LA150U 184W cross section for all other naturally occurring isotopes of tungsten. Therefore, the lack of cross-section data in the LA150U library is a definite concern when using MCNPX to determine secondary neutron production in a medical accelerator room since a majority of contamination neutrons are produced in tungsten components.