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- 저자 : Kochunas Brendan (검색결과 3 건)
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Performing linear regression with responses calculated using Monte Carlo transport codes
Dean Price,Brendan Kochunas 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.5
In many of the complex systems modeled in the field of nuclear engineering, it is often useful to uselinear regression-based analyses to analyze relationships between model parameters and responses ofinterests. In cases where the response of interest is calculated by a simulation which uses Monte Carlomethods, there will be some uncertainty in the responses. Further, the reduction of this uncertaintyincreases the time necessary to run each calculation. This paper presents some discussion on how theMonte Carlo error in the response of interest influences the error in computed linear regression coefficients. A mathematical justification is given that shows that when performing linear regression inthese scenarios, the error in regression coefficients can be largely independent of the Monte Carlo errorin each individual calculation. This condition is only true if the total number of calculations are scaled tohave a constant total time, or amount of work, for all calculations. An application with a simple pin cellmodel is used to demonstrate these observations in a practical problem.
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A multilevel in space and energy solver for multigroup diffusion eigenvalue problems
Ben C. Yee,Brendan Kochunas,Edward W. Larsen 한국원자력학회 2017 Nuclear Engineering and Technology Vol.49 No.6
In this paper, we present a new multilevel in space and energy diffusion (MSED) method for solving multigroup diffusion eigenvalue problems. The MSED method can be described as a PI scheme with three additional features: (1) a grey (one-group) diffusion equation used to efficiently converge the fission source and eigenvalue, (2) a space-dependent Wielandt shift technique used to reduce the number of PIs required, and (3) a multigrid-in-space linear solver for the linear solves required by each PI step. In MSED, the convergence of the solution of the multigroup diffusion eigenvalue problem is accelerated by performing work on lower-order equations with only one group and/or coarser spatial grids. Results from several Fourier analyses and a one-dimensional test code are provided to verify the efficiency of the MSED method and to justify the incorporation of the grey diffusion equation and the multigrid linear solver. These results highlight the potential efficiency of the MSED method as a solver for multidimensional multigroup diffusion eigenvalue problems, and they serve as a proof of principle for future work. Our ultimate goal is to implement the MSED method as an efficient solver for the twodimensional/ three-dimensional coarse mesh finite difference diffusion system in the Michigan parallel characteristics transport code. The work in this paper represents a necessary step towards that goal.
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Cross section generation for a conceptual horizontal, compact high temperature gas reactor
Kang Junsu,Seker Volkan,Ward Andrew,Jabaay Daniel,Kochunas Brendan,Downar Thomas 한국원자력학회 2024 Nuclear Engineering and Technology Vol.56 No.3
A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.
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