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Division Spotlight
Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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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.
Bin Zhang, Hongchun Wu, Yunzhao Li, Liangzhi Cao, Wei Shen
Nuclear Science and Engineering | Volume 186 | Number 2 | May 2017 | Pages 134-146
Technical Paper | doi.org/10.1080/00295639.2016.1273018
Articles are hosted by Taylor and Francis Online.
In general, spatial homogenization, energy group condensation, and angular approximation are all included in the homogenization process. For the traditional pressurized water reactor (PWR) two-step calculation, the assembly homogenization with assembly discontinuity factors plus two-group (2G) neutron diffusion calculation have been proved to be a very efficient combination. However, this changes and becomes unsettled for the pin-by-pin calculation. Thus, this paper evaluates pin-cell homogenization techniques by comparison with the two-dimensional one-step whole-core transport calculation. For the homogenization, both the generalized equivalence theory (GET) and the superhomogenization (SPH) methods are studied. Considering the spectrum interference effect between different types of fuel pin cells, both 2G and 7-group (7G) structures are condensed from the 69-group WIMS-D4 library structure. For practical reactor core applications, the low-order angular approximations, including the diffusion and the SP3 methods, are compared with each other to determine which one is accurate enough for the PWR pin-by-pin calculation. Numerical results have demonstrated that both the GET and the SPH methods work effectively in pin-cell homogenization. In consideration of the spectrum interference effect, the 7G structure is sufficient for the pin-by-pin calculation. Compared with the diffusion method, the SP3 method can decrease the errors dramatically.
