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Division Spotlight
Operations & Power
Members focus on the dissemination of knowledge and information in the area of power reactors with particular application to the production of electric power and process heat. The division sponsors meetings on the coverage of applied nuclear science and engineering as related to power plants, non-power reactors, and other nuclear facilities. It encourages and assists with the dissemination of knowledge pertinent to the safe and efficient operation of nuclear facilities through professional staff development, information exchange, and supporting the generation of viable solutions to current issues.
Meeting Spotlight
2024 ANS Winter Conference and Expo
November 17–21, 2024
Orlando, FL|Renaissance Orlando at SeaWorld
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 Science and Engineering
October 2024
Nuclear Technology
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Latest News
Don’t get boxed in: Entergy CNO Kimberly Cook-Nelson shares her journey
Kimberly Cook-Nelson
For Kimberly Cook-Nelson, the path to the nuclear industry started with a couple of refrigerator boxes and cellophane paper. Her sixth-grade science project was inspired by her father, who worked at Seabrook power station in New Hampshire as a nuclear operator.
“I had two big refrigerator boxes I taped together. I cut the ‘primary operating system’ and the ‘secondary system’ out of them. Then I used different colored cellophane paper to show the pressurized water system versus the steam versus the cold cooling water,” Cook-Nelson said. “My dad got me those little replica pellets that I could pass out to people as they were going by at my science fair.”
Donghao He, William Walters
Nuclear Science and Engineering | Volume 196 | Number 9 | September 2022 | Pages 1101-1113
Technical Paper | doi.org/10.1080/00295639.2022.2049991
Articles are hosted by Taylor and Francis Online.
The combined fission matrix (CFM) method is a newly developed neutron transport theory. This method estimates the fission matrix of the reactor core or spent fuel pool by combining a set of database fission matrices. The RAPID neutron transport code based on the CFM routine was developed originally for the spent fuel storage system and has been applied to the reactor core calculation in recent years. It can perform high-fidelity whole-core transport calculations within minutes. However, since the fission matrix database is obtained from Monte Carlo calculations, the uncertainty in the fission matrix will inevitably pass to its eigenvalue and eigenvector. The RAPID code also uses the fission matrix homogenization and interpolation techniques to further improve the calculation efficiency. Therefore, it is difficult to establish a relationship between the fission matrix elements’ uncertainty and the resulting eigenvalue and eigenvector uncertainties. This paper proposes two uncertainty analysis methods to obtain the eigenvalue and eigenvector uncertainties. The fission matrix resampling method resamples the database fission matrix elements according to each individual uncertainty. This method could generate many fission matrix databases at little additional costs and analyze the eigenvalue and eigenvector uncertainties from these resampled fission matrix coefficients. The analog uncertainty analysis method predicts the eigenvalue uncertainty from the uncertainty of the total fission rate in a fixed-source calculation, which yields a fission matrix column. Both uncertainty analysis methods have been validated against the reference brute-force calculations on a single-pin model and the BEAVRS whole-core model. It shows that the fission matrix resampling method could well estimate the uncertainties in the fission matrix eigenvalue and eigenvector. The analog uncertainty analysis method can accurately predict the eigenvalue uncertainty, which provides a guideline for the number of neutron histories simulated per fixed-source calculation.
