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Mathematics & Computation
Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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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Commercial nuclear innovation "new space" age
In early 2006, a start-up company launched a small rocket from a tiny island in the Pacific. It exploded, showering the island with debris. A year later, a second launch attempt sent a rocket to space but failed to make orbit, burning up in the atmosphere. Another year brought a third attempt—and a third failure. The following month, in September 2008, the company used the last of its funds to launch a fourth rocket. It reached orbit, making history as the first privately funded liquid-fueled rocket to do so.
Jamal Al Zain, O. El Hajjaji, T. El Bardouni, Y. Boulaich
Nuclear Science and Engineering | Volume 193 | Number 11 | November 2019 | Pages 1276-1289
Technical Paper | doi.org/10.1080/00295639.2019.1622927
Articles are hosted by Taylor and Francis Online.
This study aims to evaluate a simplified one-dimensional thermal-hydraulic module (THM) established by the DRAGON5/DONJON5 codes that allow a multiphysics study of the Syrian Miniature Neutron Source Research (MNSR) reactor both in steady-state and transient conditions. The purpose of this paper is therefore to describe the THM, fully integrated and implanted in DONJON5 to allow coupling with neutronic modules existing in the same code and to perform steady-state thermal-hydraulic and safety analyses of the reactor. Then we compare the results given by the THM with the results obtained by the Program for the Analysis of REactor Transients (PARET)/Argonne National Laboratory (ANL) thermal-hydraulic code. In order to validate our PARET/ANL and the THM in DONJON5, the fuel center temperature as a function of core power was calculated and compared with the corresponding values of the PARET code. Moreover, we have calculated the departure from nucleate boiling ratio. The comparison of the results of this study showed a good correlation between the values obtained with the THM and the thermal-hydraulic PARET/ANL code.