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Conference Spotlight
2025 ANS Winter Conference & Expo
November 8–12, 2025
Washington, DC|Washington Hilton
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 News 40 Under 40—2025
Last year, we proudly launched the inaugural Nuclear News 40 Under 40 list to shine a spotlight on the exceptional young professionals driving the nuclear sector forward as the nuclear community faces a dramatic generational shift. We weren’t sure how a second list would go over, but once again, our members resoundingly answered the call, confirming what we already knew: The nuclear community is bursting with vision, talent, and extraordinary dedication.
Jamal Al Zain, O. El Hajjaji, T. El Bardouni, M. Lahdour
Nuclear Technology | Volume 206 | Number 4 | April 2020 | Pages 620-636
Technical Paper | doi.org/10.1080/00295450.2019.1662669
Articles are hosted by Taylor and Francis Online.
The Syrian miniature neutron source reactor (MNSR), a 30-kW, 90.0% highly enriched uranium fueled (U-Al) MNSR-type reactor has gone critical. Under operating conditions of 2 h per day for 5 days a week at a peak thermal neutron flux of 1.0 × 1012 n/cm2·s, the estimated core life is 10 years. After the fuel is depleted, the full spent-fuel assembly will be replaced with new low-enriched uranium. This study presents the results of a multigroup fuel burnup and depletion analysis of the MNSR fuel lattice using the DRAGON5 transport lattice code. Furthermore, infinite multiplication factor k∞ and several two-group macroscopic parameters, including scattering cross section, fission cross section, total cross section, and diffusion coefficient, and the transport mean free path have been studied. In addition to this, fuel isotopic composition dependency on burnup was calculated as a part of this study. The results contained in this study can be used as a microscopic database for performing criticality safety analysis and shielding computations for the design of a spent-fuel storage cask for the MNSR reactor core.
