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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
2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott 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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Nuclear Science and Engineering
August 2025
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July 2025
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Latest News
Hanford proposes “decoupled” approach to remediating former chem lab
Working with the Environmental Protection Agency, the Department of Energy has revised its planned approach to remediating contaminated soil underneath the Chemical Materials Engineering Laboratory (commonly known as the 324 Building) at the Hanford Site in Washington state. The soil, which has been designated the 300-296 waste site, became contaminated as the result of a spill of highly radioactive material in the mid-1980s.
Yuanhao Gou, Conglong Jia, Zhaoyuan Liu, Kan Wang
Nuclear Science and Engineering | Volume 199 | Number 1 | April 2025 | Pages S485-S499
Research Article | doi.org/10.1080/00295639.2024.2380613
Articles are hosted by Taylor and Francis Online.
Neutron multiplicity pertains to the probability distribution of the quantity of neutrons released during induced or spontaneous fission processes within fissile materials. The technology for neutron multiplicity measurement leverages temporal correlations in the emission of fission neutrons from nuclear materials. It employs mathematical tools to elucidate the processes of neutron generation, multiplication within the nuclear material, and detection of outside nuclear materials. In this paper, two multiplicity counting methods are devised building on the RMC (Reactor Monte Carlo) code.
The results obtained from both methods, including singles, doubles, and triples counting rates, exhibit good agreement with MCNP. Additionally, parameters associated with the detection efficiency and decay time of the apparatus are computed. By amalgamating the acquired singles, doubles, and triples counting rates, the mass of fissile material within the sample is inversely determined using a passive method with the point model equation. Notably, the point model equation reveals that spontaneous fission neutrons and induced neutrons possess distinct energy spectra, challenging the validity of the assumption that the probability of neutrons being captured without causing fission can be disregarded. In light of these considerations, the neutron multiplicity counting equation was rederived. The accuracy of the Monte Carlo simulation results is improved using the new method.