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Division Spotlight
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.
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.
Milan Vujović, Miloš Vujisić
Nuclear Technology | Volume 208 | Number 11 | November 2022 | Pages 1649-1665
Technical Paper | doi.org/10.1080/00295450.2022.2070354
Articles are hosted by Taylor and Francis Online.
Several select geopolymer, polymer, and composite materials are considered as potential candidates for the inner shielding in containers used for storage and disposal of low- and intermediate-level radioactive waste, from the perspective of radiation effects. The suitability of the probed materials is examined through Monte Carlo simulations, which yield absorbed dose values in container inner shieldings of various compositions and dimensions. The radioactive waste considered in the simulation models contains 60Co or 137Cs and is placed inside standard 216.5-L (55-gal) drums, in either compacted or solidified form. The influence of container stacking, in either a storage or a disposal environment, on the dose in the shielding is also taken into account. The simulation results are used for calculating the dose-dependent overpressure within the container caused by the gas generated in the inner shielding through radiolysis. Two types of waste activity limits are determined for each of the researched shielding materials: one below which the overpressure decreases after the initial heat-induced jump and another that results in an overpressure that stays just below the maximum tolerable value. Dose-dependent changes of the polymer and composite shielding materials’ molecular weights are also calculated. The obtained results show that with regard to the radiation effects caused by the investigated sources, the examined materials are compatible with the proposed use as inner shielding in radioactive waste containers.