Multigroup Cross-Section Generation with the OpenMC Monte Carlo Particle Transport Code-Reference-Cited by-同舟云学术

Multigroup Cross-Section Generation with the OpenMC Monte Carlo Particle Transport Code

Published:2019-03-09 Issue:7 Volume:205 Page:928-944
ISSN:0029-5450
Container-title:Nuclear Technology
language:en
Short-container-title:Nuclear Technology

Author:

Boyd William¹^ORCID,Nelson Adam²,Romano Paul K.³,Shaner Samuel⁴,Forget Benoit⁴,Smith Kord⁴

Affiliation:

1. MITRE Corporation, 7525 Colshire Drive, McLean, Virginia 22102

2. Naval Reactors, 1240 Isaac Hull Avenue SE, Washington Navy Yard, District of Columbia 20376

3. Argonne National Laboratory, Mathematics and Computer Science Division, 9700 South Cass Avenue, Lemont, Illinois 60439

4. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, Building 24, Cambridge, Massachusetts 02139

Funder

Exascale Computing Project

Department of Energy, Office of Science

Idaho National Laboratory

National Science Foundation

Publisher

Informa UK Limited

Subject

Condensed Matter Physics,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Link

https://www.tandfonline.com/doi/pdf/10.1080/00295450.2019.1571828

Reference18 articles.

1. The Serpent Monte Carlo code: Status, development and applications in 2013

2. J. M. POUNDERS, “Stochastically Generated Multigroup Diffusion Coefficients,” PhD Thesis, Georgia Institute of Technology (2006).

3. OpenMC: A state-of-the-art Monte Carlo code for research and development

4. The OpenMOC method of characteristics neutral particle transport code

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