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Division Spotlight
Robotics & Remote Systems
The Mission of the Robotics and Remote Systems Division is to promote the development and application of immersive simulation, robotics, and remote systems for hazardous environments for the purpose of reducing hazardous exposure to individuals, reducing environmental hazards and reducing the cost of performing work.
National Meeting
2020 Winter Meeting and Nuclear Technology Expo
November 15–19, 2020
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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Latest News
NEA issues call to action in report on nuclear cost reductions
A new report from the Paris-based OECD Nuclear Energy Agency declares that nuclear power is needed for countries to meet their Paris Agreement decarbonization and energy security policy goals, but that governmental support for a rapid reduction in the cost of new nuclear capacity through the creation of certain policy frameworks is likely necessary.
E. Fridman, E. Shwageraus, A. Galperin
Nuclear Science and Engineering | Volume 159 | Number 1 | May 2008 | Pages 37-47
Technical Paper | dx.doi.org/10.13182/NSE07-34
Articles are hosted by Taylor and Francis Online.
Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analyzing advanced thermal and fast reactor designs for a variety of fuel compositions and geometries. The main drawback of Monte Carlo-based systems is a long calculation time imposing significant restrictions on the complexity and amount of design-oriented calculations. This paper presents an alternative approach to interfacing the Monte Carlo and depletion modules aimed at addressing this problem. The main idea is to calculate the one-group cross sections for all relevant isotopes required by the depletion module in a separate module external to Monte Carlo calculations. Thus, the Monte Carlo module will produce the criticality and neutron spectrum only, without tallying of the individual isotope reaction rates. The one-group cross section for all isotopes will be generated in a separate module by collapsing a universal multigroup (MG) cross-section library using the Monte Carlo calculated flux. Here, the term "universal" means that a single MG cross-section set will be applicable for all reactor systems and is independent of reactor characteristics such as a neutron spectrum; fuel composition; and fuel cell, assembly, and core geometries. This approach was originally proposed by Haeck et al. and implemented in the ALEPH code.Implementation of the proposed approach to Monte Carlo burnup interfacing was carried out through the BGCORE system. One-group cross sections generated by the BGCORE system were compared with those tallied directly by the MCNP code. Analysis of this comparison was carried out and led to the conclusion that in order to achieve the accuracy required for a reliable core and fuel cycle analysis, accounting for the background cross section (0) in the unresolved resonance energy region is essential.An extension of the one-group cross-section generation model was implemented and tested by tabulating and interpolating by a simplified 0 model. A significant improvement of the one-group cross-section accuracy was demonstrated.
